The current storage capacity of the spent nuclear fuel at the Kori unit 2 has reached approximately 94% saturation, excluding emergency core capacity. As South Korea has not yet constructed any interim storage facilities to store spent nuclear fuel, one of possible options is to install high density storage racks in spent fuel pool at the reactor site to increase its capacity. The high density storage rack is a technology developed to allow the storage to have more spent nuclear fuel than originally planned, while still ensuring safety. It achieves this by reducing the storage gap between the spent nuclear fuel. The purpose of this study is to investigate an appropriate storage capacity for spent fuel pool in the Kori unit 2 and the factors to be considered during the high density storage rack installation. Given that the Kori unit 2 is planning continued operation and Korea Hydro & Nuclear Power (KHNP) is preparing to install a temporary storage facility for spent nuclear fuel at the Kori nuclear site, it is reasonable to consider the installation of high density storage racks in the spent fuel pool as a storage solution for these issues. When evaluating the capacity of the spent fuel pool, the amount of spent nuclear fuel generated by other reactors in Kori nuclear site and the amount of spent nuclear fuel generated by continued operation of the Kori unit 2 should be taken into account. This study aims to consider these factors and evaluate the capacity of the spent fuel pool. Furthermore, when installing high density storage rack for the spent nuclear fuel, it should be noted that the existing storage racks at the Kori unit 2 are welded to the liner plate, which may require additional cutting work. Therefore, it is necessary to review the suitable method for the cutting work. Additionally, assuming that divers need to access the area near the storage racks or cutting & welding devices require radiation protection in the area, it is essential to analyze the expected radiation level with computational code and propose appropriate measures to limit work time or establish a work zone. Thus, this study evaluates appropriate capacity of spent fuel pool and work methods for the installation of high density storage rack in the spent fuel pool at the Kori unit 2. It is expected that this paper contributes to install high density storage racks in SFP safely.

CANDU Spent Fuel (CSF) dry storage system, SILO, has been operated from 1992 at Wolsung under 50 year operating license. As of 2023, this system has been operated for over 30 years and its licensed remaining operation time is less than 20 years. When it faces the final stage of operation, it has only two options; moving to a centralized away-from-reactor storage or extending its license atreactor. These two options have an inevitable common duty of confirming the CSF integrity by a “demonstration test”. Since the degradation of CSF and structural materials in the SILO are critically dependent on temperature, two important goals of the ‘DEMO test’ were set as follows. 1. Design of ‘DEMO SILO’: Development of internal monitoring technology by transforming SILO design. 2. Accurate measurement and evaluation of the three-dimensional temperature distribution in the ‘DEMO SILO’ Based on operating real commercial SILO dimension, a conceptual “DEMO SILO” design has been developed from 2022. Because, unlike with commercial Silo, ‘Demo Silo’ must be disassembled and assembled, and have penetration holes. Safety evaluation technologies like structural, thermal and radiation protection analysis also have been developed with design work. ‘Demo SILO’ should evaluate an accurate 3D temperature distribution with minimal number of thermocouples and penetration holes to avoid disruption of internal flow and temperature distribution. For this reason, a ‘Best Estimate Thermal-Hydraulics evaluation system for SILO’ is under development and it will be essential for ensuring temperature prediction accuracy. Construction of a full-scale test apparatus to validate this technology will begin in 2024. In order to supply power to many heaters and monitor temperature gradient inside of this apparatus, it has modular design concept by dividing its whole body to axial 9 sub-bodies which looks like a donut containing a basket at center position.

Flow-induced vibration can lead to fretting wear damage of fuel rods and spacer grids in nuclear reactors. Similarly, during the transport of spent nuclear fuel assemblies, continuous vibration and intermittent impact might also result in fretting wear due to dynamic interaction. Therefore, it is important to evaluate fuel rod damage due to fretting wear under such transport conditions. This study examines spent nuclear fuel rod specimens fabricated with hydride cladding tubes and simulated pellets, with hydrogen content ranging from 200 to 700 ppm and oxide film thickness ranging from 25 to 100 micrometers. Tests were conducted under a worst-case scenario, assuming continuous exposure to that condition during the expected transport time. Results indicate that the wear depth of all rod specimens occurred within the oxide film, suggesting a high resistance to fretting wear during transportation.

There have been a variety of issues related to spent nuclear fuel in Korea recently. Most of the issues are related to intermediate storage and disposal of spent nuclear fuel. However, recently, various studies have been started in advanced nuclear countries such as the United States to reduce spent nuclear fuel, focusing on measures to reduce spent nuclear fuel. In this study, a simple preliminary assessment of the thermal part was performed for the consolidation storage method which separates fuel rods from spent nuclear fuel and stores them. The preliminary thermal evaluation was analyzed separately for storing the spent fuel in fuel assembly state and separating the fuel rods and storing them. The consolidation storage method in separating the fuel rods was advantageous in terms of thermal conductivity. However, detailed evaluation should be performed considering heat transfer by convection and vessel shape when storing multiple fuel bundles simultaneously.

In Korea, the construction of dry storage facilities for spent nuclear fuel is being promoted through the 2nd basic plan for high-level radioactive waste management. When operating dry storage facilities, exposure dose assessment for workers should be performed, and for this, exposure scenarios based on work procedures should be derived prior. However, the dry storage method has not yet been sufficiently established in Korea, so the work procedure has not been established. Therefore, research is needed to apply it domestically based on the analysis of spent nuclear fuel management methods in major overseas leading countries. In this study, the procedure for receiving and storing spent nuclear fuel in a concrete overpack-based storage facility was analyzed. Among the various spent nuclear fuel management systems, the metal overpack-based HI-STAR 100 system and the concrete overpackbased HI-STORM 100 system are quite common methods in the United States. Therefore, in this study, work procedures were analyzed based on each final safety analysis report. First, the HI-STAR 100 overpack enters the facility and is placed in the transfer area. Remove the impact limiter of the overpack and install the alignment device on the top of the overpack. Place the HI-TRAC, an on-site transfer device, on top of the alignment unit and remove the lids of the two devices to insert the canister into the HI-TRAC. When the canister transfer is complete, reseat the lid to seal it, and disconnect the HI-TRAC from the HI-STAR 100. Raise the canister-loaded HI-TRAC over the alignment device on the top of the HI-STORM 100 overpack and remove the lids of the two devices that are in contact. Insert the canister into the HI-STORM 100 and reseat the lid. The HI-STORM 100 loaded with spent nuclear fuel is transferred to the designated storage area. In this study, the procedure for receiving and storing spent nuclear fuel in a concrete overpack-based storage facility was analyzed. The main procedure was the transfer of canisters between overpacks, and it was confirmed that HI-TRAC was used in the work procedure. The results of this study can be used as basic data for evaluating the exposure dose of operating workers for the construction of dry storage facilities in Korea.

After Fukushima nuclear power plant accident in 2011, Concerns about accident of spent fuel pool increase. In Korea, the time of saturation of spent fuel pool is coming, but regulatory measures and safety evaluation are insufficient when occurring spent fuel pool accident. Thus, it is necessary to review of spent fuel pool accident in foreign countries to establish regulatory measures and safety evaluation of spent fuel pool accident suitable for domestic spent fuel pool. Therefore, we reviewed spent fuel pool accident that occurred at Fukushima Unit 4, SONGS Unit 2 and PAKS. In Japan, spent fuel pool accident occurred at Fukushima NPP in 2011. Tsunami was cause of the accident. Station Black Out occurred at Fukushima NPP and Emergency Diesel Generator lost their functions due to Tsunami. As a result, Loss of cooling happened in spent fuel pool at Fukushima NPP. For Unit 4, wall of spent fuel pool in Unit 4 was damaged due to hydrogen explosive, so loss of coolant in spent fuel pool of Unit 4 occurred. After the accident, the temperature of spent fuel pool increases to 75°C, but there was no damage to the spent fuel. In USA, spent fuel pool accident occurred at SONGS Unit 2 in 2013. The debris of nearby ocean is cause of the accident. The debris entered the system through a damaged Salt Water Cooling pump suction strainer. The debris obstructed flow through the Component Cooling Water heat exchanger and operation of Salt Water Cooling. The maximum spent fuel pool temperature during this event was 25.6°C. It was a value that satisfied the technical specifications of the SONGS NPP. In Ukraine, spent fuel pool accident occurred at PAKS in 2003. Unintentionally opened valve of cleaning tank is cause of the accident. Loss of coolant occurred in spent fuel pool of PAKS. Due to loss of coolant, spent fuels were exposed to the vapor state atmosphere, and oxidation occurred in the cladding tube of the spent fuel that rose to 1,400°C. In this study, Review of spent fuel pool accident in major foreign countries was conducted as basic studies for establishing regulatory measures and safety evaluation of spent fuel pool in Korea. Causes of each accident were different by structure of spent fuel pools. Result of this study will be contributed to establish safety measures of spent fuel pool accident suitable for domestic spent fuel pool facility.

A radiation shielding resin with thermal stability and high radiation shielding effect has been developed for the neutron shielding resin filled in the shielding shell of dry storage/transport cask for spent nuclear fuel. Among the most commercially available neutron shielding resins, epoxy and aluminum hydroxide boron carbide are used. But in case of the resin, hydrogen content enhances the neutron shielding effect through optimization of aluminum hydroxide, zinc borate, boron carbide, and flame retardant. We developed a radiation shielding material that can increase the boron content and have thermal stability. Flame retardancy was evaluated for thermal stability, and neutron shielding evaluation was conducted in a research reactor to prove the shielding effect. As a result of the UL94 vertical burning test, a grade of V-0 was received. Therefore, it was confirmed that it had flame retardancy. According to an experiment to measure the shielding rate of the resin against neutron rays using NRF (Neutron Radiography Facility), a shielding rate of 91.54% was confirmed for the existing resin composition and a shielding rate of 96.30% for the developed resin composition. A 40 M SANS (40 M Small Angle Neutron Scattering Instrument) neutron shielding rate test was performed. Assuming aging conditions (6 hours, 180 degrees), the shielding rate was analyzed after heating. As a result of the experiment, the developed products with 99.8740% and 99.9644% showed the same or higher performance.

Molten Salt Reactor (MSR) is one of Generation-IV nuclear reactors that uses molten salts as a fuel and coolant in liquid forms at high temperatures. The advantages of MSR, such as safety, economic feasibility, and scalability, are attributed from the fact that the molten salt fuel in a liquid state is chemically stable and has excellent thermo-physical properties. MSR combines the fuel and coolant by dissolving the actinides (U, Th, TRU, etc.) in the molten salt coolant, eliminating the possibility of a core meltdown accident due to loss of coolant (LOCA). Even if the molten salt fuel leaks, the radioactive fission products dissolved in the molten salt will solidify with the fuel salt at room temperature, preventing potential leakage to the outside. MSR was first demonstrated at ORNL starting with the Aircraft Reactor Experiment (ARE) in 1954 and was extended to the 7.4 MWth MSRE developed in 1964 and operated for 5 years. Recently, various start-ups, including TerraPower, Terrestrial Energy, Moltex Energy, and Seaborg, have been conducting research and development on various types of MSR, particularly focusing on its inherent safety and simplicity. While in the past, fluoride-based molten salt fuels were used for thermal neutron reactors, recently, a chlorine-based molten salt fuel with a relatively high solubility for actinides and advantageous for the transmutation of spent nuclear fuel and online reprocessing has been developing for fast neutron spectrum MSRs. This paper describes the development status of the process and equipment for producing highpurity UCl3, a fuel material for the chlorine-based molten salt fuel, and the development status of the gas fission product capturing technologies to remove the gaseous fission products generated during MSR operation. In addition, the results of the corrosion property evaluation of structural materials using a natural circulation molten salt loop will also be included.

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.

In this paper, we report and discuss the semi-permanently hydrophilic (SPH) treatment of polyester fabric using plasma polymerization and oxidation based on atmospheric pressure dielectric barrier discharge (APDBD) technology. SiOxCy (-H) was coated on polyester fabric using Hexamethylcyclotrisiloxane (HMCTSO) as a precursor, and then plasma oxidation was performed to change the upper layer of the thin film to SiO2-like. The degradation of hydrophilicity of the SPH polyester fabrics was evaluated by water contact angle (WCA) and wicking time after repeated washing. The surface morphology of the coated yarns was observed with scanning electron microscopy, and the presence of the coating layer was confirmed by measuring the Si peak using energy dispersive x-ray spectroscopy. The WCA of the SPH polyester fabric increased to 50 degrees after 30 washes, but it was still hydrophilic compared to the untreated fabric. The decrease in hydrophilicity of the SPH fabric was due to peeling of the SiOxCy(-H) thin film coated on polyester yarns.