With an ultimate view to identifying abnormal releases of radioactive materials, a set of liquid and gaseous effluent data including unplanned or uncontrolled releases annually reported form the U.S. and Korean nuclear power plants were systematically analyzed. With the use of 21 years’ worth of annual discharge data for 7 radionuclide groups and 24 individual radionuclides, taken from a combined total of 1,610 reactor-years (RYs) covering 62 units of US Pressurized Water Reactors (PWRs) and 22 units of Korean PWRs, three novel formulas for estimating events were employed to calculate characteristic values. Applying these characteristic values derived from the event estimation formulas to events that transpired during 699 RYs in operational US PWRs revealed an enhanced predictive accuracy for abnormal events when considering individual radionuclides, as opposed to grouping them by radionuclide groups. This effect was particularly pronounced for specific events such as leaks caused by problems in Gas Decay Tanks, leaks in Steam Generator Power Operated Relief Valves, fuel defects, and leaks during spent nuclear fuel processing. In the case of Korean PWRs, fuel defects were identified as the primary events related to radioactive effluent releases. The methodologies and characteristic values derived from this study were applied to these events. The event estimation rate was lower in Korean compared to US PWRs, which can be attributed to the lower frequency of event occurrences in Korean PWRs (30 RYs) compared to the US. The approach proposed in this study may contribute to develop a methodology to identify implicit abnormal release data and correlate them with specific operational occurrences or events, which could improve the conventional practice of simply recording and reporting radioactive discharge data.

The Korea Research Institute of Standards and Science has developed certified reference materials (concrete, soil, and metal radioactive liquid) for measuring gamma-emitting radionuclides to improve and maintain the quality assurance and quality control of the radioactivity measurement in decommissioning nuclear power plants. The raw materials that make up each CRM were mixed in an appropriate ratio with radionuclides. For certification and homogeneity assessment, 10 bottles were randomly selected, two sub-samples were collected from each bottle, and radionuclides were measured via HPGe gamma spectrometry. The results of the homogeneity tests using a one-way analysis of variance on the radionuclides in the CRMs fulfilled the requirements of ISO Guide 35. Coincidence summing and self-absorption correction were performed on measurement results by introducing the Monte Carlo efficiency transfer code and Monte Carlo N-Particle transport code. In concrete analysis, the reference values for five radionuclides (60Co, 241Am, 134Cs, and 137Cs) in the CRM were in the range of 15-40 Bq/kg, and the expanded uncertainty was within 10% (k = 2). In soil analysis, the reference values for the 137Cs and 60Co were 118.7 and 124.4 Bq/kg, and the expanded uncertainty was within 10% (k = 2). In metal radioactive liquid analysis, the reference values for 134Cs, 137Cs and 60Co in the CRM were in the range of 200-270 Bq/kg, and the expanded uncertainty was within 7% (k = 2).

Domestic commercial low- and intermediate-level radioactive waste storage containers are manufactured using 1.2 mm thick cold-rolled steel sheets, and the outer surface is coated with a thin layer of primer of 10~36 μm. However, the outer surface of the primer of the container may be damaged due to physical friction, such as acceleration, resonance, and vibration during transportation. As a result, exposed steel surfaces undergo accelerated corrosion, reducing the overall durability of the container. The integrity of storage containers is directly related to the safety of workers. Therefore, the development of storage containers with enhanced durability is necessary. This paper provides an analysis of mechanical properties related to the durability of WC (tungsten carbide)-based coating materials for developing low- and intermediate-level radioactive waste storage containers. Three different WC-based coating specimens with varied composition ratios were prepared using HVOF (high-velocity oxy-fuel) technique. These different specimens (namely WC-85, WC-73, and WC-66) were uniformly deposited on cold-rolled steel surfaces ensuring a constant thickness of 250 μm. In this work, the mechanical properties of the three different WCbased coaitng materials evaluated from the viewpoints of microstructure, hardness, adheision force between substrate and coating material, and wear resistance. The cross-sectional SEM-EDS (Scanning Electron Microscope-Energy Dispersive X-ray Spectroscopy) images revealed that elements W (tungsten), C (carbon), Ni (nickel), and Cr (chromium) were uniformly distributed within the each coating layers which was approximately 250 μm thick. The average hardness values of HWC-85 and HWC-73 were found to be 1,091 Hv (Vickers Hardness) and 1,083 Hv, respectively, while the HWC-66 exhibited relatively lower hardness value of 883 Hv. This indicates that a higher WC content results in increased hardness. Adhesion force between and substrates and coating materials exceeded 60 MPa for all specimens, however, there were no significant differences observed based on the tungsten carbide content. Furthermore, a taber-type abrasion tester was used for conducting abrasion resistance tests under specific conditions including an H-18 load weight at 1,000 g with rotational speed set at 60 RPM. The abrasion resistance tests were performed under ambient temperatures (RT: 23±2°C) as well as relative humidity levels (RH: 50±10%). Currently, the ongoing abrasion resistance tests will include some results in this study.

This study focuses on the development of coatings designed for storage containers used in the management of radioactive waste. The primary objective is to enhance the shielding performance of these containers against either gamma or neutron radiation. Shielding against these types of radiation is essential to ensure the safety of personnel and the environment. In this study, tungsten and boron cabide coating specimens were manufactured using the HVOF (High-Velocity Oxy Fuel) technuqe. These coatings act as an additional layer of protection for the storage containers, effectively absorbing and attenuating gamma and neutron radiation. The fabricated tungsten and boron carbide coating specimens were evaluated using two different testing methods. The first experiment evaluates the effectiveness of a radiation shielding coating on cold-rolled steel surfaces, achieved by applying a mixture of WC (Tungsten Carbide) powders. WC-based coating specimens, featuring different ratios, were prepared and preliminarily assessed for their radiation shielding capabilities. In the gamma-ray shielding test, Cs-137 was utilized as the radiation source. The coating thickness remained constant at 250 μm. Based on the test results, the attenuation ratio and shielding rate for each coated specimen were calculated. It was observed that the gammaray shielding rate exhibited relatively higher shielding performance as the WC content increased. This observation aligns with our findings from the gamma-ray shielding test and underscores the potential benefits of increasing the tungsten content in the coating. In the second experiment, a neutron shielding material was created by applying a 100 μm-thick layer of B4C (Boron Carbide) onto 316SS. The thermal neutron (AmBe) shielding test results demonstrated an approximate shielding rate of 27%. The thermal neutron shielding rate was confirmed to exceed 99.9% in the 1.5 cm thick SiC+B4C bulk plate. This indicates a significant reduction in required volume. This study establishes that these coatings enhance the gamma-ray and neutron shielding effectiveness of storage containers designed for managing radioactive waste. In the future, we plan to conduct a comparative evaluation of the radiation shielding properties to optimize the coating conditions and ensure optimal shielding effectiveness.

Properties of bentonite, mainly used as buffer and/or backfill materials, will evolve with time due to thermo-hydro-mechanical-chemical (THMC) processes, which could deteriorate the long-term integrity of the engineered barrier system. In particular, degradation of the backfill in the evolution processes makes it impossible to sufficiently perform the safety functions assigned to prevent groundwater infiltration and retard radionuclide transport. To phenomenologically understand the performance degradation to be caused by evolution, it is essential to conduct the demonstration test for backfill material under the deep geological disposal environment. Accordingly, in this paper, we suggest types of tests and items to be measured for identifying the performance evolution of backfill for the Deep Geological Repository (DGR) in Korea, based on the review results on the performance assessment methodology conducted for the operating license application in Finland. Some of insights derived from reviewing the Finnish case are as follows: 1) The THMC evolution characteristics of backfill material are mainly originated from hydro-mechanical and/or hydrochemical processes driven by the groundwater behavior. 2) These evolutions could occur immediately upon installation of backfill materials and vary depending on characteristics of backfill and groundwater. 3) Through the demonstration experiments with various scales, the hydro-mechanical evolution (e.g. advection and mechanical erosion) of the backfill due to changes in hydraulic behavior could be identified. 4) The hydro-chemical evolution (e.g. alteration and microbial activity) could be identified by analyzing the fully-saturated backfill after completing the experiment. Given the findings, it is judged that the following studies should be first conducted for the candidate backfill materials of the domestic DGR. a) Lab-scale experiment: Measurement for dry density and swelling pressure due to saturation of various backfill materials, time required to reach full saturation, and change in hydraulic conductivity with injection pressure. b) Pilot-scale experiment: Measurement for the mass loss due to erosion; Investigation on the fracture (piping channel) forming and resealing in the saturation process; Identification of the hydro-mechanical evolution with the test scale. c) Post-experiment dismantling analysis for saturated backfill: Measurement of dry density, and contents of organic and harmful substances; Investigation of water content distribution and homogenization of density differences; Identification of the hydro-chemical evolution with groundwater conditions. The results of this study could be directly used to establishing the experimental plan for verifying performance of backfill materials of DGR in Korea, provided that the domestic data such as facility design and site characteristics (including information on groundwater) are acquired.

Engineered Barrier Systems (EBS) are a key element of deep geological repositories (DGR) and play an important role in safely isolating radioactive materials from the ecosystem. In the environment of a DGR, gases can be generated due to several factors, including canister corrosion. If the gas production rate exceeds the diffusion rate, pore pressures may increase, potentially inducing structural deterioration that impairs the function of the buffer material. Therefore, understanding the hydraulic-mechanical behavior of EBS due to gas generation is essential for evaluating the longterm stability of DGR. This study employed X-ray computed tomography (CT) technology to observe cracks created inside the buffer material after laboratory-scale gas injection experiments. After CT scanning, we identified cracks more clearly using an image analysis method based on machine learning techniques, enabling us to examine internal crack patterns caused by gas injection. In the samples observed in this study, no cracks were observed penetrating the entire buffer block, and it was confirmed that most cracks were created through the radial surface of the block. This is similar to the results observed in the LASGIT field experiment in which the paths of the gas migration were observed through the interface between the container and the buffer material. This study confirmed the applicability of high-resolution X-ray CT imaging and image analysis techniques for qualitative analysis of internal crack patterns and cracks generated by gas breakthrough phenomena. This is expected to be used as basic data and crack analysis techniques in future research to understand gas migration in the buffer material.

In the nuclear environment, sensors ensure safety, monitoring, and operational efficiency under various operating conditions. These sensors come in various forms, each tailored to specific purposes, including nuclear safety and security, waste treatment and storage, gas leak detection, temperature and humidity monitoring, and corrosion detection. Ensuring the longevity of sensors without the need for frequent replacements is a vital goal for researchers in this field. This paper explores materials that can act as shields to protect sensors from harsh environmental conditions (high radiation and temperatures) to enhance their lifetime. The types of material that had been explored were divided into categories: metal and non-metal. Fourteen types of metal and seven different plastic materials were studied and focused on their characteristics and current applications. Considering properties like melting point, intensity, and conductivity, plastic materials are chosen to be examined as sensor shielding material. A preliminary experiment was conducted to verify signal characteristics changes by shielding material. Metal material and plastic material each were placed in the middle of the granite and the target sensor. The result showed that when metal is between the granite and the sensor, the density and impedance are higher in granite than in the metal. This leads to signal attenuation and a shift in resonance frequency, while plastic does not. Therefore, PPS (Polyphenylene sulfide) and PAI (Polyamide-imide) have lower density and impedance than granite while also possessing heat, moisture, and radiation resistance for effective shielding.

The compacted bentonite buffer is a key component of the engineered barrier system in deep geological repositories for high-level radioactive waste disposal. Groundwater infiltration into the deep geological repository leads to the saturation of the bentonite buffer. Bentonite saturation results in bentonite swelling, gelation and intrusion into the nearby rock discontinuities within the excavation damaged zone of the adjacent rock mass. Groundwater flow can result in the erosion and transport of bentonite colloids, resulting in bentonite mass loss which can negatively impact the long-term integrity and safety of the overall engineered barrier system. The hydro -mechanicalchemical interactions between the buffer, surrounding host rock and groundwater influence the erosion characteristics of the bentonite buffer. Hence, assessing the critical hydro-mechanicalchemical factors that negatively affect bentonite erosion is crucial for the safety design of the deep geological repository. In this study, the effects of initial bentonite density, aperture, discontinuity angle and groundwater chemistry on the erosion characteristics of Bentonil WRK are investigated via bentonite extrusion and artificial fracture experiments. Both experiments examine bentonite swelling and intrusion into simulated rock discontinuities; cylindrical holes for bentonite extrusion experiments and plane surfaces for artificial fracture experiments. Compacted bentonite blocks and bentonite pellets are manufactured using a compaction press and granulation compactor respectively and installed in the transparent extrusion cells and artificial fracture cells. The reference test condition is set to be 1.6 g/cm3 dry density and saturation using distilled water. After distilled water or solution injection, the axial and radial expansion of the bentonite specimens into the simulated rock discontinuities are monitored for one month under free swelling conditions with no groundwater flow. Subsequent flow tests are conducted using the artificial fracture cell to determine the critical flow rate for bentonite erosion. The intrusion and erosion characteristics are modelled using a modified hydro-mechanicalchemical coupled dynamic bentonite diffusion model and a fluid-based hydro-mechanical penetration model.

Currently, the most promising fuel candidate for use in sodium fast reactors (SFRs) is metallic fuel, which is produced by a modified casting method in which the metallic fuel material is sequentially melted in an inert atmosphere to prevent volatilization, followed by melting in a graphite crucible, and then injection casting in a quartz (SiO2) mold to produce metallic fuel slugs. In previous studies, U-Zr metallic fuel slugs have been cast using Y2O3 reaction prevent coatings. However, U-Zr alloy-based metallic fuel slugs containing highly reactive rare earth (RE) elements are highly reactive with Y2O3-coated quartz (SiO2) molds and form a significant thickness of surface reaction layer on the surface of the metallic fuel slug. Cast parts that have reacted with nuclear fuel materials become radioactive waste. To decrease amount of radioactive waste, advanced reaction prevent material was developed. Each RE (Nd, Ce, Ln, Pr) element was placed on the reaction prevent material and thermal cycling experiments were carried out. In casting experiments with U-10wt% Zr, it was reported that Y2O3 layer has a high reaction prevent performance. Therefore, the reaction layer properties for RE elements with higher reactivity than uranium elements were evaluated. To investigate the reaction layer between RE and NdYO3, the reaction composition and phase properties as a function of RE content and location were investigated using SEM, EDS, and XRD. The results showed that NdYO3 ceramics had better antireaction performance than Y2O3.

Ring Tensile Test (RTT) is mainly performed for comparing tensile strength and total strain between nuclear fuel cladding specimens under various initial conditions. Through RTT, the loaddisplacement (F-D) curve obtained from the uniaxial tensile test can also be obtained. However, the Young’s modulus estimated from the gradient of the straight portion is much lower than general value of materials. The reasons include tensile machine compliance, slack in the fixtures, or elastic deformation of the fixtures and the tooling. Another reason is that the bending of the test part in the ring is stretched with two pieces of tools. Although the absolute value of the Young’s modulus is smaller than the actual value, it is applicable to calculate the ratio of the Young’s moduli of different materials, that is, the relative value. The Young’s modulus, or slope of the linear section, varies slightly depending on which location data is used and how much data is included. In order to obtain a more accurate ratio of Young’s moduli between materials using the RTT results, a post-processing method for the ring tensile test results that can prevent such human errors is proposed as follows. First, the slope of the linear section is obtained using the displacement and load when the load increase is the largest and the displacement and load of the position that is 95% of the maximum load increase. To replace the section where the ring-shaped specimen is stretched at the beginning of the F-D curve, a straight line equal to the slope of the linear section is drawn to the displacement axis from the position of maximum load increase and moved to the origin to obtain the final F-D curve for a RTT. Lastly, the yield stress uses the stress at the point where the 0.2% offset straight line and the F-D curve meet as suggested in the ASTM E8/E8M-11 “Standard test methods for tensile testing of metallic materials”. RTT results post-processing method was coded using FORTRAN language so that it could be performed automatically. In addition, sensitivity analysis of the included data range on the Young’s modulus was performed by using the included data range as 90%, 85%, and 80% of the maximum load increase.

Molten salt reactor (MSR) uses fluoride or chloride based molten salt as a coolant of the system, and fuel materials are dissolved in the molten salt, therefore it can be act as both coolant and nuclear fuel. A few issues have arisen from early-stage research and development program of MSR from Oak Ridge National Laboratory, including corrosion of structural materials and fission product management. For investigating the effect of additives on corrosion of structural materials, Mg(OH)2 and MgCl2*6H2O are added into the NaCl-MgCl2 eutectic salt. Prepared chloride salt is injected into the autoclave in the glove box, as well as corrosion coupons for candidate structural materials for molten chloride salt reactor, SS316, Alloy 600, and C-276 are also prepared. The temperature is set as 700°C. After 500 h corrosion experiment, the samples are taken out from the autoclave, and they are analyzed with scanning electron microscopy (SEM) and energy-dispersive X-ray spectroscopy (EDS). SS316 samples show weight loss with all salt conditions, while Alloy 600 and C-276 show weight gain after the corrosion experiment.

Korea has signed nuclear cooperation agreements (NCA) with 29 countries. Nuclear materials, materials, equipment, and technology transferred under the agreements are “internationally controlled materials (ICM)” under the Nuclear Safety Act. The main obligations imposed on those items include ensuring peaceful use, safeguards, physical protection, annual inventory reporting, and retransfer with supplier prior consent. The Nuclear Export and Import Control System (NEPS) handles the export control procedures for transferring ICM. After import, inventory management for ICM in Korea would be transitioning to an item-based system through the Obligation Tracking System for internationally controlled item (OTS) currently under development. A one-stop import and export control system for ICM can be established when information is well-linked between these two systems. This paper aims to derive a methodology for integrating NEPS and OTS. NEPS-OTS coupling begins at the receipt confirmation and shipment notification stages in NEPS. When importing ICM under NCA, the inventory change (code RF: receipt foreign) is entered in OTS by getting the information that has completed the receipt confirmation in NEPS. Conversely, during export, the information that has completed the shipment notification procedure in NEPS is linked to the OTS so that the entire cycle from import to re-export of the ICM can be concluded. Inventory verification for retransfer, checking that the book inventory remains positive value, is impossible under the current system. This issue can be resolved by enabling inventory information in OTS to be displayed in NEPS. Determining when and how to generate the obligation code for imported ICM is essential for NEPS-OTS coupling. Manual input may be necessary for some cases with multiple obligations. Nevertheless, it is more efficient from a system communication protocol to automatically generate and impose a single obligation based on the supplier country information in NEPS. Moreover, it is important to automatically link crucial information available in NEPS to reduce the administrative burden on OTS users and discrepancies between systems. Most required OTS data, such as country obligation, item categories, quantity, physical or chemical form, and receipt date, can be directly linked from NEPS. However, NEPS improvement is needed for digitizing the receiver information and facility data, like the material balance area. The NEPS-OTS integration involves sharing data as a system and encompasses the connection between export control and inventory management. Future work to link some information in NEPS -OTS with the KSIS could be suggested to enhance efficiency and effectiveness in managing ICM.

In this study, we successfully grafted chitosan (CS) onto multi-walled carbon nanotubes (MWCNTs) to enhance their properties and potential applications in the biomedical field. FTIR spectroscopy confirmed the successful covalent bonding of CS onto MWCNTs, indicated by the new absorption peak of the amide bond (–CONH–). Thermal analysis showed that the modified MWCNTs (MWCNT-CS) had significant weight loss around 260 °C, suggesting the decomposition of hydroxypropyl chitosan, and confirming its presence in the nanocomposite. SEM images revealed that CS grafting improved the dispersibility of MWCNTs, a property crucial for their use as nanofillers in polymers. Moreover, the micro-tensile bond strength of dentin surface increased with increasing MWCNT-CS concentrations, indicating the potential of MWCNT-CS as a pretreatment for dentin bonding. After simulated aging, the bond strength remained significantly higher for MWCNT-CS groups compared to those without pretreatment. In biocompatibility assessment using the MTT assay, MWCNT-CS showed higher cell viability than MWCNT, suggesting improved biocompatibility after CS modification. The results of this study suggest that CS-modified MWCNTs could be promising materials for applications in dentin bonding, dentin mineralization, bone scaffolding, implants, and drug delivery systems.

Proton exchange membrane fuel cells (PEMFCs) are an auspicious energy conversion technology with the potential to address rising energy demands while reducing greenhouse gas emissions. The stack’s performance, durability, and economy scale are greatly influenced by the materials used for the PEMFC, viz., the membrane electrocatalyst assembly (MEA) and bipolar flow plates (BPPs). Despite extensive study, carbon-based materials have outstanding physicochemical, electrical, and structural attributes crucial to stack performance, making them an excellent choice for PEMFC manufacturers. Carbon materials substantially impact the cost, performance, and durability of PEMFCs since they are prevalently sought for and widely employed in the construction of BPPs and gas diffusion layers (GDLs)) and in electrocatalysts as a support material. Consequently, it is essential to assemble a review that centers on utilizing such material potential, focusing on its research development, applications, problems, and future possibilities. The prime focus of this assessment is to offer a clear understanding of the potential roles of carbon and its allotropes in PEMFC applications. Consequently, this article comprehensively evaluates the applicability, functionality, recent advancements, and ambiguous concerns associated with carbonbased materials in PEMFCs.

시설 재배 시, 미세먼지의 잦은 발생은 피복재의 광투과율을 감소시키고 이는 작물의 생육에 간접적인 영 향을 미칠 수 있다. 본 연구에서는 미세먼지 발생에 따른 폴리에틸렌(PE)과 폴리올레핀(PO) 필름의 광투과율 변화 를 조사하고, 피복재의 광투과율 감소에 따른 봄철 재배 오이의 생육 변화를 확인하였다. 미세먼지 발생 챔버를 이 용하여 PE와 PO 필름을 지속적으로 미세먼지에 노출시켰을 때, PE 필름에서 미세먼지 발생에 의한 광투과율 감소 가 PO 필름보다 크게 나타났다. PE 필름에 인위적으로 먼지를 부착시켜서 대조구 대비 10, 20 및 30% 광투과율 감 소 처리구를 설정한 후 오이를 재배하였을 때, 3월 말 이후 재배 후반기에서의 오이 생육은 광투과율 감소 처리구에 서 증가하였으나, 누적 수확량은 대조구에서 가장 높았다. 봄철 오이 재배에서 미세먼지 발생에 의한 광투과율 감 소는 3월 말 이후 시설 내 고온 노출에 의한 생육 지연을 줄일 수 있었으나, 전 생육 기간 동안의 광투과율 감소는 입 사광량의 감소 및 광합성의 저하로 오이의 총 수확량을 감소시켰다.

이 연구는 핵무기 폭발 시 발생하는 효과 변인을 토대로 북한이 언제, 어떤 방법으로 핵무기를 운용할 것이며, 핵폭발 시 생성되는 방사성 물 질이 자연환경과 인공물의 영향에 따라 도심지에서 어떤 거동 현상을 보 이는가와 이를 고려한 국민 방호의 대비 방향에 관한 것이다. 연구 결과 핵무기는 폭발 고도에 따라 그 효과가 달라지며, 북한은 이를 활용하여 개전 초부터 가장 효과적인 공격을 할 것으로 예측되었다. 즉, 북한은 개 전 초 한미연합군과 정부의 지휘‧통제‧통신체계를 무력화하기 위해 저위 력핵무기로 지하 폭발을, 전쟁 도중 결정적인 목표 확보를 위해 전술핵 무기로 저공 폭발을, 전쟁 말기 패색이 짙어지는 위기 시에는 전술핵무 기로 지표면 폭발을 시도할 것이다. 북한의 핵무기 공격 후 발생되는 방 사성 물질은 낙진의 형태로 일정 지역을 오염시킬 것이며, 방사성 물질 이 도심지로 유입된다면 공기역학 또는 유체역학적 거동을 보임으로써 다양한 형태의 오염과 위험이 존재할 것으로 분석되었다. 이에 따라 국 민 방호를 위해서는 북한의 핵무기 공격 양상을 고려 최악의 상황을 가 정한 대비가 평시에 완료되어야 하며, 전쟁 개시 이후에는 당시의 공격 유형에 부합한 대응 및 복구가 뒤따라야 한다. 아울러 방사능 낙진의 거 동을 세밀히 분석하고 이를 고려하여 핵폭발 초기 효과에 대비하는 주민 대피와 이를 후속하는 낙진에 대응하기 위한 주민 소개는 분리되어야 한다.

This study focused on improving the phase stability and mechanical properties of yttria-stabilized zirconia (YSZ), commonly utilized in gas turbine engine thermal barrier coatings, by incorporating Gd2O3, Er2O3, and TiO2. The addition of 3-valent rare earth elements to YSZ can reduce thermal conductivity and enhance phase stability while adding the 4-valent element TiO2 can improve phase stability and mechanical properties. Sintered specimens were prepared with hot-press equipment. Phase analysis was conducted with X-ray diffraction (XRD), and mechanical properties were assessed with Vickers hardness equipment. The research results revealed that, except for Z10YGE10T, most compositions predominantly exhibited the t-phase. Increasing the content of 3-valent rare earth oxides resulted in a decrease in the monoclinic phase and an increase in the tetragonal phase. In addition, the t(400) angle decreased while the t(004) angle increased. The addition of 10 mol% of 3-valent rare-earth oxides discarded the t-phase and led to the complete development of the c-phase. Adding 10 mol% TiO2 increased hardness than YSZ.

Transition metal chalcogenides are promising cathode materials for next-generation battery systems, particularly sodium-ion batteries. Ni3Co6S8-pitch-derived carbon composite microspheres with a yolk-shell structure (Ni3Co6S8@C-YS) were synthesized through a three-step process: spray pyrolysis, pitch coating, and post-heat treatment process. Ni3Co6S8@C-YS exhibited an impressive reversible capacity of 525.2 mA h g-1 at a current density of 0.5 A g-1 over 50 cycles when employed as an anode material for sodium-ion batteries. However, Ni3Co6S8 yolk shell nanopowder (Ni3Co6S8-YS) without pitch-derived carbon demonstrated a continuous decrease in capacity during charging and discharging. The superior sodium-ion storage properties of Ni3Co6S8@C-YS were attributed to the pitchderived carbon, which effectively adjusted the size and distribution of nanocrystals. The carbon-coated yolk-shell microspheres proposed here hold potential for various metal chalcogenide compounds and can be applied to various fields, including the energy storage field.

상변화 물질(PCM)은 상전이 동안 에너지를 흡수하거나 방출할 수 있는 잠열 저장 물질로 활용된다. 최근 수십 년 동 안, 연구자들은 다양한 온도 적용을 위한 건설 물질로의 다양한 PCM의 통합을 탐구해 왔다. 그러나, PCM을 통합하는 콘크리트 의 기계적 및 열적 반응은 통합 방법에 의해 영향을 받는다. PCM을 콘크리트에 추가하기 위한 여러 기술이 제안되었다. 그럼 에도 불구하고, 콘크리트에 마이크로 캡슐화 PCM(m-PCM)의 통합은 종종 기계적 강도의 상당한 감소를 초래한다. 기존 콘크리 트에 m-PCM의 추가와 관련된 한계를 극복하기 위해, 예외적인 강도 및 내구성 특성으로 인해 초고성능 시멘트 복합체(UHPCC) 가 선호된다. 따라서, 본 연구에서는 기존 기술의 단점을 해결하기 위해 PCM을 통합한 신규 나노 엔지니어링 UHPCC를 개발하 였다. 또한, 시멘트 복합체의 기계적 및 열적 성능을 향상시키기 위해 다중 벽 탄소 나노튜브(MWCNT)를 추가하였다. 결과는 MWCNT의 포함이 기계적 성능을 향상시켰을 뿐만 아니라 시멘트 복합체의 열적 성능을 향상시켰다는 것을 보여 주었다.