In this study, radioactivity of Cs-134, Cs-137, and Eu-154, which are gamma-emitting nuclides among fission products of spent fuel, was analyzed as a tool to measure the burnup of spent fuel nondestructively. This nuclide has a unique gamma-ray energy, allowing the amount of the isotope to be estimated based on the intensity of the gamma-ray at a specific energy. The SCALE 6.2 ORIGAMI (ORIGen AsseMbly Isotopics) module and the latest ORIGEN-arp library were used for computational analysis. The spent fuel samples were selected as WH14×14 with an enrichment of 1.5~5.0wt%, a burnup of 10~60 GWD/MTU, and a cooling time of 0~40 years. The analysis results were benchmarked using SFCOMPO experimental data provided by OECD/ NEA, including isotope inventory and uncertainty measured by destructive radiochemical analysis, fuel assembly design data required for benchmark model development, reactor design information, and operating history information. 16 similar spent fuels were selected from SFCOMPO data and the calculation results of Cs-134, Cs-137, and Eu-154 were compared. The average error of the Cs-134 radioactivity calculation result was 2.81%, and the maximum error was 6.70%. The average errors of Cs-137 and Eu-154 were 2.42% and 4.95%, respectively, and the maximum errors were 5.40% and 14.91%, respectively.

Spent Fuel Pool Island (SFPI) is a spent nuclear fuel storage pool that operates independently of existing nuclear facilities to safely manage SNF and minimize maintenance costs during the nuclear decommissioning process. Since the radiation controlled area can be dismantled before transporting SNF to a dry storage facility, the overall decommissioning period can be shortened, and the risk of occupational exposure during dismantling is reduced. In the US, various nuclear power plants have introduced SFPI for this reason. In this paper, to analyze the economic feasibility of application of SFPI to nuclear power plants to be decommissioned, several scenarios are established in consideration of the decommissioning plan and schedule, SFPI and dry storage facility application schedule. Cost and benefit list (SFPI application cost, SNF management cost, SNF dry storage cask cost, etc.) for each alternative were derived, and economic analysis was conducted by applying the Net Present Value (NPV). As a result of the analysis, it is found that the application of SFPI during decommissioning is economically effective as the NPV showed a positive number even when uncertainty was taken into account.

In case of damaged spent fuels, it would require additional treatment for their transportation and storage to capture the radioactive fission products in a defined space. The canning container for the damaged spent fuels is one way to seal the radioactive fission products inside the container. In the Post Irradiation Examination Facility (PIEF) of KAERI, the Quiver container has been introduced for canning damaged spent fuels from Westinghouse Sweden. The main container body has been manufactured for particle-tightness of spent fuel. In addition, drying equipment is being prepared for gas-tightness of spent fuel. The drying equipment can remove water and fill the inert gas inside the container. Before drying inside the container, we evaluated the volatile fission products inventory because volatile fission products could be released during the drying process. Despite assuming highly conservative hypotheses for the inventory remaining in damaged fuel rods, the amount that could be released during the drying process was less and dose rate levels around the evacuation piping system were low.

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.

After spent fuel is stored in a dry storage container, it becomes difficult to obtain information on the fuel’s characteristics. As a result, it is necessary to identify the characteristics of spent nuclear fuel in advance and secure the information necessary to establish delivery acceptance requirements for interim storage and disposal in the future. Therefore, it is necessary to evaluate the characteristics of spent fuel before loading dry storage casks. In order to prepare for the dry storage of spent fuel, information on the basic characteristics of the fuel is required. As part of this information, it is also necessary to establish calculation criteria for spent fuel burnup. Spent fuel burnup can be classified into three categories. The first is burnup evaluated using design codes (design burnup), the second is burnup measured by furnace instruments during power plant operation (actual burnup), and the third is burnup measured through measurement equipment (measured burnup). This paper describes a comparative evaluation of design burnup, actual burnup, and measured burnup for specific fuels (40 bundles).

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.

Spent fuel from the Wolsong CANDU reactor has been stored in above-ground dry storage canisters. Wolsong concrete dry storage canisters (silos) are around 6 m high, 3 m in outside diameter, and have shielding comprised of around 1 m of concrete and 10 mm of steel liner. The storage configuration is such that a number of fuel bundles are placed inside a cylindrical steel container known as a Fuel Basket. The canisters hold up to 9 baskets each that are 304 L stainless steel, around 42” in diameter, 22” in height, and hold 60 fuel bundles each. The operating license for the dry storage canisters needs to be extended. It is desired to perform in-situ inspections of the fuel baskets to very their condition is suitable for retrieval (if necessary) and that the temperature within the fuel baskets is as predicted in the canister’s design basis. KHNP-CNL (Canadian Nuclear Lab.) has set-up the design requirements to perform the in-situ inspections in the dry storage canisters. This Design Requirements applies to the design of the dry storage canister inspection system.

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.

The DPRK had been upgrading its nuclear weapons capabilities from the past to the sixth nuclear test in 2017, and Kim Jong-un has been in power since the death of Kim Jong-il in 2011, striving to upgrade and diversify four nuclear tests and firing means. In 2022, in particular, DPRK launched more than 40 ballistic and cruise missiles and provoked them in various ways, such as developing solid rocket engines, flying fighter jets, and invading drones. In addition, reprocessing facility activities have been observed again since 2021 at the Yongbyon Nuclear Research Complex. Operational activities such as continuous activities of the 5MWe Yongbyon reactor and the additional construction of new buildings are observed. DPRK’s recent activities could result in nuclear weapons in all except conventional weapons provocations. DPRK has researched and developed nuclear weapons since the 1950s. It has been preparing to operate nuclear weapons, operating nuclear power, and modernizing nuclear power simultaneously. Given the number of nuclear weapons using DPRK’s nuclear materials and various means of missiles, an offensive transition is expected to enable restrictive deterrence strategies that can be used first use and on assured retaliation. In addition, based on the nuclear strategy, which is interpreted as Vipin Narang’s nuclear doctrine and nuclear posture, DPRK is also capable of assured retaliation and asymmetric escalation posture. In particular, considering the continuous activities of the Yongbyon Nuclear Research Complex, which has recently diversified the investment vehicle, and the delegation of nuclear weapon use, it is expected to move differently from the previous one based on the changed nuclear strategy. However, there are clear limitations to interpreting it as a completely assured retaliation and asymmetric expansion nuclear strategy. First, there is a lack of development of atmospheric reentry vehicles that can avoid ICBM interception for assured retaliation capabilities against the United States. Second, there are limitations in the open capacity of nuclear weapons due to the absence of SSBN capabilities. However, delegation to operations at strategy force suggests the possibility of asymmetric expansion strategies. The previous analysis of DPRK’s nuclear strategy and limitations is valid in that the U.S. nuclear umbrella guarantees the Republic of Korea in a strong alliance between the Republic of Korea and the U.S. If the Republic of Korea lacks a nuclear umbrella due to the weakening of the alliance or limits U.S. intervention by having more than dozens of ICBMs, it is considered that DPRK can use a definite confirmation retaliation and asymmetric escalation nuclear strategy. As a response to this, it is the first way to verify and strengthen the validity of the three-axis system (Kill Chain, KAMD, KMPR), second to strengthen the Korea-U.S. alliance, and finally to appeal to the international community and increase consensus. In particular, it is possible to form a consensus of sanctions and condemnation DPRK by expressing concerns about nuclear dominoes caused by nuclear proliferation and arms competition to the international community.

With the introduction and implementation of the National Research and Development Innovation Act in 2021, researchers are required to have a greater understanding of research ethics and to comply more strictly. The range of misconduct in research and the standards for sanctions have been expanded with the introduction of the National Research and Development Innovation Act. In addition, researchperforming institutions and specialized agencies have been obligated to establish their own research management systems and standards according to the changed criteria. The Korea Institute of Nuclear Nonproliferation and Control (KINAC), a nuclear regulatory authority that is conducting national R&D in related fields, has sought to strengthen research ethics by revising related regulations, introducing a plagiarism detection system, and expanding related education in accordance with these policies. In this study, we analyzed the effectiveness of the plagiarism detection system as a basic quality control measure for research results and a tool for enhancing research ethics, which was introduced. KINAC did not simply introduce a plagiarism detection program but established institutional improvements and other regulatory measures to support it, with the aim of more effectively managing research results. To analyze the effectiveness of this system, we calculated the plagiarism rate by sampling 30 papers each year for the three years before the introduction of the plagiarism detection system. When comparing the plagiarism rates before and after the introduction of the plagiarism detection system, no exceptional cases of high plagiarism rates were found in papers published after the introduction of the system. Although most of the papers before the introduction of the system showed a satisfactory plagiarism rate, some cases showed high plagiarism rates. We analyzed the cause of such cases in detail. Some exceptional cases were also found to be included in the range of misconduct regulated by the National Research and Development Innovation Act. As no such cases were found after the introduction of the system, we could infer that the system is effectively functioning as a tool for basic quality control and enhancing research ethics. In the future, we plan to expand the sample qualitatively and quantitatively by including other forms of outcomes published by the institution, not just papers, and conduct a more detailed analysis. Based on the results, we will develop various improvement plans for enhancing the quality and research ethics of the institution’s research results.

Both ISO 21001 and ISO 9001 are standards developed by the International Organization for Standardization (ISO) for quality management systems. However, while ISO 9001 focuses on the general requirements for Quality Management Systems (QMS), ISO 21001 is specifically designed for educational institutions. ISO 9001 is a widely recognized standard for QMS applicable to almost industries, including manufacturing and services. It defines the requirements for establishing, implementing, maintaining, and continuing improvement of QMS to improve customer satisfaction by meeting customer requirements and improving overall performance. Meanwhile, ISO 21001 focuses specifically on educational institutions and is designed to develop and improve the curriculum efficiency by meeting trainees needs. It provides a system in which educational institutions can build, implement, maintain, and continuously improve the Education Management System (EMS) for the purpose of improving the satisfaction of trainees and other stakeholders. ISO 21001 covers a wide range of educational organizations, including schools, universities, and education providers. KINAC/INSA, the Center of Excellence in Korea, is an educational institution in the field of nuclear control. So It has been developed and operated various international and domestic curriculum. KINAC/INSA obtained ISO 9001 certification in November 2016 and has been certified so far. However, in the scope of ISO 9001 certification, curriculum development process is not included so KINAC/ INSA needs to obtain additional ISO certification specialized in education to improve the education quality. That is why KINAC/INSA is developing the ISO 21001 system, and aims to acquire certification in November 2023. This paper explains the necessity for educational institutions to obtain ISO 21001 in comparison with ISO 9001. It also introduces the process of developing ISO 21001 system of KINAC/INSA. By implementing EMS based on ISO 21001, KINAC/INSA can expect to improve the educational satisfaction of trainees and other stakeholders through effective curriculum development and educational operation.

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.

빠르게 변화하는 해외 시장 환경 및 소비자 니즈의 변화에 발맞춰, 기업의 혁신 전략의 중요성이 강조되고 있음에도 불구하고, 경쟁전략과 신제품개발속의 관계에 관한 연구가 많이 이루어지지 않 은 실정이다. 특히, 대기업에 비해 유무형 자원의 부족에 한계를 가지고 있는 중소기업에게 있어, 해외 시장 환경의 변화에 대처하는 경쟁전략의 선택은 중요한 의사결정 요소이며, 이러한 경쟁전략 은 신제품개발속도에 영향을 미치는 주요한 전략적 요인으로 작용한다. 이에 본 연구는, 차별화전 략과 비용우위전략이 신제품개발속도에 미치는 영향을 알아보고, 시장역동성이 이러한 관계에 어 떠한 조절효과를 미치는지 살펴보고자 한다. 이를 위해 한국 중소수출업체를 대상으로 설문을 진행 하였으며, 차별화전략과 비용우위전략이 신제품개발속도에 긍정적인 영향을 미치는 것을 확인하였 다. 나아가, 시장역동성이 높을수록, 차별화전략이 신제품개발속도에 미치는 긍정적인 영향은 커지 지만, 비용우위전략이 신제품개발속도에 미치는 긍정적인 영향은 약해지는 것을 확인하였다. 본 연 구는, 경쟁전략이 신제품개발속도에 미치는 영향을 제안하고 검증하였으며, 시장이 빠르게 변화할 경우, 경쟁전략의 긍정적인 효과가 달라질 수 있음을 확인하였다는 점에서 학문적 및 실무적 시사 점이 있을 것으로 기대된다.

Micro-electronic gas sensor devices were developed for the detection of carbon monoxide (CO), nitrogen oxides (NOx), ammonia (NH3), and formaldehyde (HCHO), as well as binary mixed-gas systems. Four gas sensing materials for different target gases, Pd-SnO2 for CO, In2O3 for NOx, Ru-WO3 for NH3, and SnO2-ZnO for HCHO, were synthesized using a sol-gel method, and sensor devices were then fabricated using a micro sensor platform. The gas sensing behavior and sensor response to the gas mixture were examined for six mixed gas systems using the experimental data in MEMS gas sensor arrays in sole gases and their mixtures. The gas sensing behavior with the mixed gas system suggests that specific adsorption and selective activation of the adsorption sites might occur in gas mixtures, and allow selectivity for the adsorption of a particular gas. The careful pattern recognition of sensing data obtained by the sensor array made it possible to distinguish a gas species from a gas mixture and to measure its concentration.

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.