Governments around the world are enacting laws mandating explainable traceability when using AI(Artificial Intelligence) to solve real-world problems. HAI(Human-Centric Artificial Intelligence) is an approach that induces human decision-making through Human-AI collaboration. This research presents a case study that implements the Human-AI collaboration to achieve explainable traceability in governmental data analysis. The Human-AI collaboration explored in this study performs AI inferences for generating labels, followed by AI interpretation to make results more explainable and traceable. The study utilized an example dataset from the Ministry of Oceans and Fisheries to reproduce the Human-AI collaboration process used in actual policy-making, in which the Ministry of Science and ICT utilized R&D PIE(R&D Platform for Investment and Evaluation) to build a government investment portfolio.

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.

In this study, we have fabricated the phenolic resin (PR)/polyacrylonitrile (PAN) blend-derived core-sheath nanostructured carbon nanofibers (CNFs) via one-pot solution electrospinning. The obtained core-sheath nanostructured carbon nanofibers were further treated by mixed salt activation process to develop the activated porous CNFs (CNF-A). Compared to pure PAN-based CNFs, the activated PR/PAN blend with PR 20% (CNF28-A)-derived core-sheath nanostructured CNFs showed enhanced specific capacitance of ~ 223 F g− 1 under a three-electrode configuration. Besides, the assembled symmetric CNF28-A//CNF28-A device possessed a specific capacitance of 76.7 F g− 1 at a current density of 1 A g− 1 and exhibited good stability of 111% after 5,000 galvanostatic charge/discharge (GCD) cycles, which verifies the outstanding long-term cycle stability of the device. Moreover, the fabricated supercapacitor device delivered an energy density of 8.63 Wh kg− 1 at a power density of 450 W kg− 1.

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%.

Kori unit 1, the first PWR (Pressurized Water Reactor) in Korea, was permanent shut down in 2017. 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, the types depending on generation characteristics and radiological characterization methods and process of solid radioactive waste were analyzed. Solid radioactive waste depending on the characteristics of the generation was classified into reactor vessel and reactor vessel internal, large components, small metals, spent nuclear fuel storage racks, insulation, wires, concrete debris, scattering concrete, asbestos, mixed waste, soil, spent resins and filters, and dry active waste. Radiological characterization of solid radioactive waste is performed to determine the characteristics of radioactive contamination, including the type and concentration of radionuclides. It is necessary to ensure the representativeness of the sample for the structures, systems and components to be evaluated and to apply appropriate evaluation methods and procedures according to the structure, material and type of contamination. Therefore, the radiological characterization is divided into concrete and structures, systems and components, and reactor vessel, reactor vessel internal and bioshield concrete. In this study, the types depending on generation characteristics and radiological characterization methods and process of solid radioactive waste were analyzed. The results of this study can be used as a basis for the preparation of the FDP for the Kori unit 1.

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.

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.

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.

The nuclear facilities at Korea Atomic Energy Research Institute (KAERI) have generated a variety of liquid radioactive waste and most of them have low-level radioactive or lower levels. Some of the liquid radioactive waste generated in KAERI is transported to Radioactive Waste Treatment Facility (RWTF) in 20 L container. Liquid radioactive waste transported in a 20 L container is stored in a Sewer Tank after passing through a solid-liquid separation filter. It is then transferred to a very low-level liquid radioactive waste Tank after removing impurities such as sludge through a pre-treatment device. The previous pre-treatment process involved an underwater pump and a cartridge filter device passively, but this presented challenges such as the inconvenience of having to install the underwater pump each time, radiation exposure for workers due to frequent replacement of the cartridge filter, and the generation of large amounts of radioactive waste from the filter. To address these challenges and improve efficiency and safety in radiation work, an automated liquid radioactive waste pre-treatment device was developed. The automated liquid radioactive waste pre-treatment device is a pressure filtration system that utilizes multiple overlapping filter plates and pump pressure to effectively remove impurities such as sludge from liquid radioactive waste. With just the push of a button, the device automatically supplies and processes the waste, reducing radiation hazards and ensuring worker safety. Its modular and mobile design allows for flexible utilization in various locations, enabling efficient pre-treatment of liquid radioactive waste. To evaluate the performance of the newly constructed automated liquid radioactive waste treatment device, samples were taken before and after treatment for 1 hour cycling and analyzed for turbidity. The results showed that the turbidity after treatment was more than about four times lower than before treatment, confirming the excellent performance of the device. Also, it is expected that the treatment efficiency will improve further as the treatment time and number of cycles increase.

South Korea has been storing UNF in spent fuel pool dry storage facility within Nuclear Power Plants. The dry storage facility of used nuclear fuel (UNF) is essential to sustain safety and sustain stable operation of a nuclear power plant. Most abroad countries have attempted to develop a variety of dry storage facility for used nuclear fuel in order to retain the safe restoration. Many studies have been conducting to safety evaluation for the dry storage facility. However, there is not a ventilation evaluation in the wake of fire event that could influence of the thermal effect on the dry storage facility, even though it will likely to occur fire events such as wildfire, air craft crash. In practice, it happened to catastrophic disaster due to the wild fire adjacent to ul-jin mountain. Also, it happened to fire accident near to the Japonia NPP in Ukraine territory caused of military air plane missile. It has not mostly been studied on the ventilation evaluation considered to thermal safety in the dry storage facility excepted for some researches. It could need the mechanical ventilation systems such as HVAC system in the dry storage system, so that thermal effect can be reduced. In this study, we conducted to the ventilation control modelling by using fire modelling tool (Fire Dynamic Simulator v.6.7). The ventilation scenarios made up for 3 case that can compare flowrate variation with ventilation control. As a result of modelling, there is no differentiation between ventilation control using performance curve with not using performance curve even though the pressure fluctuation would be increased, compared with the case of considering performance curve. Second, it evaluated that the mode for fraction control would occur to pressure rise in the state of controlling the ventilation system flowrate. However, sensitivity of flowrate control was more decreased below less than 5 seconds. Third, in the case of on/off control system revealed more higher resolution than other cases caused by flowrate variation. These results could be considered as the design guidelines for the development dry storage facility to improve the thermal performance that can reduce thermal risk. Furthermore, the study results would expect HVAC system installed in dry storage to help automatic ventilation control relevant to dry storage safety increased.

To evaluate the safeguards system or performance in a facility, it is crucial to analyze the diversion path for nuclear materials. However, diversion paths can range from the extremely simplified to the complicated depending on the level of knowledge and the specific person conducting the analysis. This study developed the diversion path analysis tools using an event tree and fault tree method to generating diversion paths systematically. The essential components of the diversion path were reviewed, and a logical flow was developed for systematically creating the diversion path. An algorithm was created based on the facility design components and logical flow, as well as the initial information of the nuclear materials and material flows. The event tree and fault tree analysis tools were used to test the path generation algorithm. The usage and limitations of these two logic methods are discussed, and ideas to incorporate the logic algorithm into practical program tools are suggested. The tests were analyzed on a typical light water reactor as an example, including automatic generation of dedicated pathways, configuration of safeguards measures, and analyzing paths with strategies for avoiding safeguard systems. The results led to the development of a draft pathway analyzer program that can be applied to general nuclear systems. The results of this study will be used to develop a program module that can systematically generate diversion paths using the event tree and fault tree method. It can help to guide and provide practical tools for implementing SBD.

Background: Individuals with scapular winging have a weak serratus anterior (SA) muscle, and to compensate, the pectoralis major (PM) and upper trapezius (UT) muscles excessively activate, which can cause upper extremity dysfunction. This study aimed to compare the effects of isometric horizontal abduction (IHA) on SA, PM, and UT muscle activity, as well as the SA/PM and SA/UT muscle activity ratios during knee push-up plus (KPP) at 90° and 120° of shoulder flexion. Objects: This study aimed to compare the effects of IHA on SA, PM, and UT muscle activity, as well as the SA/PM and SA/UT muscle activity ratios during KPP at 90° and 120° of shoulder flexion. Methods: This study, conducted at a university research laboratory, included 20 individuals with scapular winging. Participants performed KPP with and without IHA at 90° (KPP90) and 120° (KPP120) of shoulder flexion. SA, PM, and UT muscle activity were measured using surface electromyography. Results: PM activity in KPP90 with IHA was significantly lower than KPP90 and in KPP120 was significantly lower than KPP90. UT activity was significantly greater with IHA than without IHA and at 120° than 90° of shoulder flexion. SA/PM muscle activity ratio was significantly higher in KPP90 with IHA than without IHA and in KPP120 than in KPP90. SA/UT muscle activity ratio was significantly lower with IHA than without IHA. Conclusion: KPP90 with IHA and KPP120 are effective exercises to reduce PM activity and increase SA/PM muscle activity ratio. However, applying IHA in KPP90 also reduces SA/UT muscle activity ratio, implying that it would be preferable to apply KPP120 in individuals overusing their UT muscles.