This study explores the development and characterization of hard carbon anodes for sodium-ion batteries produced from waste coffee grounds, synthesized at both 1000 °C and 1500 °C. Importantly, this work highlights the potential of using biomass-derived hard carbons as sustainable and effective material for anode for sodium-ion batteries, contributing to the advancement of energy storage systems with increasing global demands for environmentally friendly and cost-effective technologies. The research focuses on the electrochemical performance of these hard carbons, examining how different carbonization temperatures impact their structural and electrochemical properties. Utilizing advanced analytical methods, the structural changes correlating with temperature increase were identified, including modifications in carbon atom arrangements, which significantly influence the electrochemical behaviors of the hard carbons. Our research specifically focuses on how the structural differences affect the division of capacity contribution from sloping region (above 0.1 V) and plateau regions (below 0.1 V). Electrochemical test results revealed that hard carbon with higher degree of order and reduced microstructural defects, demonstrated improved capacity values. At the same time, the highly ordered hard carbon exhibits drastic capacity loss upon increasing of current densities. The results from this study not only advance our understanding of hard carbons but also open pathways for the future exploration of hard carbons for additional improvements.

Fossil fuels have a high energy density, meaning they contain a significant amount of energy per unit of volume, making them efficient for energy production and transport. Biodiesel is especially becoming a fossil fuel alternative and a key part of renewable energy. Several types of waste from homes, markets, street vendors, and other industrial places were collected and transesterified with Ni-doped ZnO nanoparticles for this study. These included castor oil, coffee grounds, eggshells, vegetable oil, fruit peels, and soybean oil. The Ni-doped ZnO’s were then calcined at 800 °C. The maximum conversion rate found in converting fruit peel waste into biodiesel is about 87.6%, and it was 89.6% when the oil-to-methanal ratio was about 1:2 and the reaction time was 140 min. This is the maximum biodiesel production compared to other wastes. Moreover, using vegetable oil with nanocatalyst, the maximum biodiesel production rate of about 90.58% was recorded with 15% catalyst loading, which is the maximum biodiesel production compared with the other wastes with nanocatalyst. Furthermore, at 75 °C and a concentration of catalyst of about 15% the maximum biodiesel production obtained by using castor oil is about 92.8%. It has the highest biodiesel yield compared with the yield recorded from other waste. The catalyst also demonstrated great stability and reusability for the synthesis of biodiesel. Using waste fruit peels with Ni-doped ZnO helps to progress low-cost and ecologically friendly catalyst for sustainable biodiesel production.

This study addresses the environmental impact associated with waste management and natural aggregate production. It explores the potential of utilizing Coal Bottom Ash (CBA) and Reclaimed Asphalt Pavement Aggregate (RAPA) as complete replacements, respectively, for fine and coarse aggregates in concrete. Despite their similarities to natural aggregates, CBA and RAPA often end up in landfills. Laboratory tests were conducted, revealing satisfactory performance in drying shrinkage and air void parameters. However, while the flexural strength met design requirements, the compressive and splitting tensile strengths were lower than predicted. The deviation in strength development behavior from natural aggregate concrete (NAC) was attributed to weak agglomerated aggregates in RAPA and the large size of the interfacial transition zone (ITZ) due to the old asphalt coating surrounding RAPA. To enhance the strength behavior, two methods were employed: compaction in the form of roller-compacted concrete and RAPA abrasion carried out by rolling RAPA in a concrete mixer. Compaction improved aggregate interlock, while RAPA abrasion decreased agglomerated aggregates and minimized asphalt coating, reducing ITZ size. These treatments resulted in improvements in compressive, flexural, and splitting tensile strengths, with the combination of both treatments having the most significant effect. Analysis of relationships between flexural, splitting tensile, and compressive strengths indicated that CBA and RAPA concrete behaved more similarly to NAC after the treatments. This research suggests that with appropriate interventions, it is feasible to utilize CBA and RAPA in concrete, contributing to sustainable construction through improved waste management, carbon footprint reduction, and conservation of natural resources.

The waste secondary battery contains a significant amount of valuable metals, making its recycling highly desirable. However, conventional chemical methods for recycling are environmentally unfriendly and cost-ineffective. Rather than the chemical method, this paper deals with a mechanical method for recovering electrode materials from waste secondary batteries by blowing pressurized air onto the interface area between the electrode and the separator. Especially, in this study, the effective blowing angle were searched by simulating the separation of the electrode material from the separator through 1-way fluid structure interaction analysis based on the Cohesive Zone Modeling technique.

이 글은 남한 동해안에 유입된 북한 생활쓰레기로부터 시작된다. 그동안 북한 쓰레기가 남한에 유입된다는 사실은 해양쓰레기 실태조사 관련 연구 에서 주로 이루어졌다. 환동해권 해양쓰레기 유입 등에 관한 연구에서도 주로 한국과 일본, 극동러시아 등의 유입 현황에 관한 연구는 있지만, 북 한은 논의의 대상에서 제외되었다. 본 연구는 기존의 해양학 관점이 아닌 북한학 관점에서 남한 해안에 유입된 북한 쓰레기 문제를 다루었다. 북한 생활쓰레기 중 상품포장지는 직접적으로 북한 상품 생산 현황과 브랜드 등을 알 수 있으며, 간접적으로는 북한 내 경제 상황과 상품 유통 지역망 등을 알 수 있기 때문이다. 동해안 지역에서 수거한 북한제품 포장지를 살 펴보면 대부분 생산공장은 평양으로 표기되었다. 이를 통해 평양에서 생산 한 제품이 동해안 지역으로 유통됨을 알 수 있다. 또한 동해안 지역은 북 한을 대표하는 대도시인 원산, 청진, 함흥, 라선 등이 있는데, 실제로 상품 포장지에는 이 지역 생산공장이 표기되는 사례도 있었다. 김정은 집권 이 후 매년 국가적 차원에서 국가디자인전시회를 개최할 만큼 산업미술을 강 조하는데, 특히 상품의 고유한 특징을 표현하는 상표도안을 강조한다. 본 연구에서는 동해안 주요 도시에서 생산한 제품을 중심으로 같은 품목이지 만 공장별로 어떻게 상표도안이 다른지 살펴봤다. 북한 쓰레기에 대한 북 한학적 시각과 해양학적 시각의 학제간 연구를 통해 남한에 유입되는 북 한 생활쓰레기에 대한 연구의 폭을 넓혀갈 필요가 있다.

The carbon anode material for lithium-ion battery was prepared by pyrolysis fuel oil and waste polyethylene terephthalate (PET) additive. The pitch was synthesized as a medium material for carbon anode by heat treatment. The waste PET additive improved the softening point and thermal stability of the pitch. La and Lc of the anode material (heat-treated pitch) increased at higher treatment temperature but decreased by waste PET additive. The electric capacity was evaluated based on effects of defective cavity and developed graphite interlayer, respectively. When the La and Lc of the anode material decreased, the electric capacity by cavity increased based on defective graphite structure. Therefore, the addition of waste PET causes the improved capacity by the cavity. The anode material which has a high efficiency (over 95%) and C-rate (95%, 2 C/0.1 C) was obtained by controlling the process of heat treatment and PET addition. The mechanism of lithium-ion insertion was discussed based on effects of defective cavity and developed graphite interlayer.

Sustainable biomass-derived porous carbons demonstrate excellent capacitive properties owing to their heteroatom-rich nature and distinct textural feature. Herein, a series of nitrogen-/phosphorus-/oxygen-containing microporous carbons (CWWN/ P/O-MPCs) have been successfully fabricated by etching in H2O2 solution, pre-treatment of camphor wood wastes with KOH solution and subsequent carbonization. As an electrode material for supercapacitors, the typical microporous carbon (CWW-N/P/O-MPCs-0.5) exhibits a remarkably high specific capacitance of 245 F g− 1 at 0.5 A g− 1, corresponding to an impressively large volumetric capacitance of 208 F m− 3, and excellent long-term stability over 10,000 cycles. The excellent electrochemical performance can be ascribed to the optimal combination of heteroatom groups and ultrafine micropores.

온실단지 내에서 발생되는 폐영농자재의 효율적인 관리방안을 제시하기 위한 목적으로 시설농업용 폐영농자재의 관리 실태를 조사하여 검토하였다. 조사 결과를 요약하면 다음과 같다. 현재 국내에서 발생되는 영농폐기물은 아직 생활쓰레기와 같이 취급되고 있었고, 일부 영농폐기물에 대한 자료는 별도로 관리되고 있었지만 공급량에 대한 정보가 전혀 없는 등 다소 신뢰도가 떨어지는 것을 알 수 있었다. 농촌지역에 보급되는 영농자재의 경우, 개인이 필요에 따라 수시로 구입하는 등 관리 및 조사가 용이하지는 않지만, 지역농협과 농업기술관리센터 등을 적극 활용하여 영농자재의 공급에서부터 수거, 재활용 및 처리단계까지 투명하게 관리할 수 있는 시스템의 도입이 절실히 필요한 것으로 나타났다. 그리고 영농자재도 재활용이 가능한 친환경 자재의 개발도 적극 도입하여야 할 것으로 판단되었다. 특히 공동 집하장의 경우, 재정지원으로 그칠 것이 아니라 재정지원은 물론 작목반, 마을 단위 또는 지자체별로 영농폐기물에 대한 인식을 지역 주민들을 대상으로 적극적으로 홍보 및 교육 등을 실시하여 지역주민이 그 지역의 환경지킴이 역할을 할 수 있도록 하여야 할 것으로 판단되었다.

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.

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.

In the design of HLW repositories, it is important to confirm the performance and safety of buffer materials at high temperatures. Most existing models for predicting hydraulic conductivity of bentonite buffer materials have been derived using the results of tests conducted below 100°C. However, they cannot be applied to temperatures above 100°C. This study suggests a prediction model for the hydraulic conductivity of bentonite buffer materials, valid at temperatures between 100°C and 125°C, based on different test results and values reported in literature. Among several factors, dry density and temperature were the most relevant to hydraulic conductivity and were used as important independent variables for the prediction model. The effect of temperature, which positively correlates with hydraulic conductivity, was greater than that of dry density, which negatively correlates with hydraulic conductivity. Finally, to enhance the prediction accuracy, a new parameter reflecting the effect of dry density and temperature was proposed and included in the final prediction model. Compared to the existing model, the predicted result of the final suggested model was closer to the measured values.

Radioactive waste generated during decommissioning of nuclear power plants is classified according to the degree of radioactivity, of which concrete and soil are reclassified, some are discharged, and the rest is recycled. However, the management cost of large amounts of concrete and soil accounts for about 40% of the total waste management cost. In this study, a material that absorbs methyl iodine, a radioactive gas generated from nuclear power plants, was developed by materializing these concrete and soil, and performance evaluation was conducted. A ceramic filter was manufactured by forming and sintering mixed materials using waste concrete, waste soil, and by-products generated in steel mills, and TEDA was attached to the ceramic filter by 5wt% to 20wt% before adsorption performance test. During the deposition process, TEDA was vaporized at 95°C and attached to a ceramic filter, and the amount of TEDA deposition was analyzed using ICP-MS. The adsorption performance test device set experimental conditions based on ASTM-D3808. High purity nitrogen gas, nitrogen gas and methyl iodine mixed gas were used, the supply amount of methyl iodine was 1.75 ppm, the flow rate of gas was 12 m/min, and the supply of water was determined using the vapor pressure value of 30°C and the ideal gas equation to maintain 95%. Gas from the gas collector was sampled to analyze the removal efficiency of methyl iodine, and the amount of methyl iodine detected was measured using a methyl iodine detection tube.

Lubricant oil waste contaminated with radioactive materials generated at nuclear facilities can be disposed of as industrial waste in accordance with self-disposal standards if only radioactive materials are removed. Lubricant oil used in nuclear facilities consists of oil of 75-85% and additives of 15-25%, and lubricant oil waste contains heavy metals, carbon, glycol, etc. In addition, lubricant oil waste from nuclear facilities contains metallic gamma-ray emission radionuclides including Co-60, Cs-137 and volatile beta-ray emission radionuclides such as C-14 and H-3, which are not present in lubricant oil waste from general industries and these radionuclides must be eliminated according to the Atomic Energy Act. In general industries, the wet treatment technologies such as acid-white soil treatment, ion purification, thin film distillation, high temperature pyrolysis, etc. are used as the refining technology of lubricant oil waste, but it is difficult to apply these technologies to nuclear industrial sites due to restrictions related with controlling the generation of secondary radioactive waste in sludge condition containing radionuclides of metal components, and limiting the concentration of volatile radioactive elements contained in refined oil to be below the legal threshold. In view of these characteristics, the refinement system capable of efficiently refining and treating lubricant oil waste contaminated with radioactive materials generated in nuclear facilities has been developed. The treatment process of this R&D system is as follows. First, the moisture in the radioactive lubricant oil waste pretreated through the preprocessing system is removed by the heated evaporating system, and the beta-emission radionuclides of H-3 and C-14 can be easily removed in this process. Second, the heated lubricant oil waste by the heated evaporating system is cooled through the heat exchanging system. Third, the particulate matters with gamma-ray emission radionuclides are removed through the electrostatic ionizing system. Forth, the lubricant oil waste is stored in the storage tank and the purified lubricant oil waste is discharged to the outside after sampling and checking from the upper, middle and lower positions of the lubricant oil waste stored in the storage tank. Using this R&D system, it is expected that the amount of radioactive waste can be reduced by efficiently refining and treating lubricant oil waste in the form of organic compounds contaminated with radioactive materials generated in nuclear facilities.

Backfill is one of the main components of engineered barrier in a high-level waste repository. The material selection of the backfill determines the barrier performance of the backfill. Overseas, its related research has been carried out mainly in Sweden, Finland, Canada, and Japan. However, Korea has recently started backfill research, and it is urgent to select a potential material for establishing the concept of backfill material and conducting backfill research. This study reviews NEA report, potential materials for overseas backfill research, advantages and disadvantages of single and mixed backfill materials, cases of license applications in Finland and Sweden for the selection of potential materials for backfill in Korea’s high-level waste repository. The review results indicated that it is reasonable to carry out backfill research according to the following plan: Both single and mixed materials are considered as potential materials for backfill research; experiments and performance studies are conducted with these materials; and, based on the results, a potential material or candidate material for the backfill suitable for the HLW repository in Korea is determined. For this plan, the single material is tentatively selected, as in Sweden, as bentonite with a montmorillonite content of about 40-50%. Then, if the selection criteria for montmorillonite content are determined through experiments and performance studies, we determine the final potential backfill material. As for the mixed backfill material, the bentonite/crushed rock mixture seems to be more advantageous than the bentonite/sand mixture considering the disposing problem of crushed rock generated from tunnel excavation and economic feasibility through its recycling. It is thought that the bentonite used in the bentonite/crushed rock mixture should have a higher montmorillonite content than bentonite used as a single backfill material since the crushed rock acts as an inert material in the mixture. The results of this study can be used as basic data for selecting the backfill material to be applied to the high-level waste repository in Korea, and can be used as a guideline for selecting the potential material required for backfill experiments and performance studies to be carried out in the future.

As the design life of nuclear power plants are coming to the end, starting with Kori unit 1, nuclear power related organizations have been actively conducted research on the treatment of nuclear power plant decommissioning waste. In this study, among various types of radioactive waste, stabilization and volume reduction experiments were conducted on radioactive contaminated soil waste. Korea has no experience in decommissioning nuclear power plants, but a large amount of radioactively contaminated soil waste was generated during the decommissioning of the KAERI research reactor (TRIGA Mark- II) and the uranium conversion facility. This case shows the possibility of generating radioactive soil waste from nuclear power plants and nuclear-related facilities sites. Soil waste should be solidified, because its fluidity and dispersibility wastes specified in the notification of the Korea Nuclear Safety and Security Commission. In addition, the solidified waste forms should have sufficient mechanical strength and water resistance. Numerous minerals in the soil are components that can make glass and ceramics, for this reason, glass-ceramic sintered body can be made by appropriate heat and pressure. The sintering conditions of soil were optimized, in order to make better economical and more stable sintered body, some additives (such as additives for glass were mixed) with the soil and sintering experiments were conducted. Uncontaminated natural soil was collected and used for the experiment after air drying. Moisture content, pH, bulk density, and organic content were measured to understand the basic properties of soil, and physicochemical properties of the soil were identified by XRD, XRF, TG, and SEM-EDS analysis. In order to understand the distribution by particle size of the soil, it was divided into Sand (0.05–2 mm) and Fines (< 0.05 mm). The green body was manufactured in the form of a cylinder with a diameter of 13mm and a height of about 10mm. Appropriate pressure (> 150 MPa) was applied to the soil to make a green body, and appropriate heat (> 800°C) was applied to the sintered body to make a sintered body. The sintering was conducted in a muffle furnace in air conditions. The volume reduction and compressive strength of the sintered body for each condition were evaluated.

Radioactive Cesium is fission products of spent nuclear fuelwith high heat generating nuclide, having a 30 years half-life. Particularly, it is important to make stable waste form because Cs-137 have high solubility and mobility at ground water. The ceramic waste form has higher thermal and structural stability and lower solubility than glass and cement waste form. Various ceramic waste forms for Cs immobilization have been researched such as aluminosilicate (CsAlSi2O6), phosphate (CsZr2(PO4)3), titanate (CsxAlxTi8-XO16) and CsZr0.4W1.5O6. Cs pollucite is incorporated radio-Cesium to aluminosilicate framework by inorganic ion-exchange with zeolite. Therefore, it is an extremely stable structure. In previous study, we are prepared Cs pollucite pellet with various ratio of Cs precursor/matrix materials, and attempted to evaluate applicability as ceramic waste form. Cs pollucite is produced by mixing Mullite and SiO2 obtained by heat treatment Kaolinite with Cs2CO3 in ratios of 0.5, 0.6, 0.7, 0.8. Optimized ratio was 0.5 revealed single pollucite phase and the others exhibited CsAlSiO4 phase with pollucite. Cs pollucite of ratio 0.5 was pelletized under various conditions and evaluated performance as waste form. herein, the pellets were cracked on surface and edges broken. Therefore, Cs pollucite having high ratio of matrix materials contained Si and Al was prepared and pelletized, and then waste form was evaluated. The Cs pollucite powder is ratio of Cs precursor/matrix materials were 0.1, 0.2, 0.3, 0.4. Pollucite powder was mixed with 1.5, 2.0wt% Polyvinyl alcohol as binder, and dried at 70°C for overnight. Afterward, these powders obtained were pressed using punch-die apparatus at 50, 100 bar for 1 hour and the pellets with about dia. 25 mm and height 10 mm was acquired. These pellets were sintered at 1,400°C for 5 hours. Subsequently, the waste forms were evaluated physicochemical test such as compression strength, thermal conductivity, thermal expansion and leaching properties analysis.