• Proceedings of the Korean Radioactive Waste Society Conference
• /
• 2005.11a
• /
• pp.175-176
• /
• 2005

• Proceedings of the Korean Radioactive Waste Society Conference
• /
• 2017.05a
• /
• pp.279-280
• /
• 2017

• Proceedings of the Korean Radioactive Waste Society Conference
• /
• 2011.10a
• /
• pp.123-124
• /
• 2011

• Nuclear Engineering and Technology
• /
• v.43 no.2
• /
• pp.149-158
• /
• 2011

Metal waste generated from the pyroprocessing of 10 MtU of spent fuel was classified by comparing the specific activity of a relevant radionuclide with the limit value of the specific activity specified in the Korean acceptance criteria for a lowand intermediate-level waste repository. A Korean Optimized Fuel Assembly design with a 17${\times}$17 array, an initial enrichment of 4.5 weight-percent, discharge burn-up of 55 GWD/MtU, and a 10-year cooling time was considered. Initially, the mass and volume of each structural component of the assembly were calculated in detail, and a source term analysis was subsequently performed using ORIGEN-S for these components. An activation cross-section library generated by the KENO-VI/ORIGEN-S module was utilized for top-end and bottom-end pieces. As a result, an Inconel grid plate, a SUS plenum spring, a SUS guide tube subpart, SUS top-end and bottom-end pieces, and an Inconel top-end leaf spring were determined to be unacceptable for the Gyeongju low- and intermediate-level waste repository, as these waste products exceeded the acceptance criteria. In contrast, a Zircaloy grid plate and guide tube can be placed in the Gyeongju repository. Non-contaminated Zircaloy cladding occupying 76% of the metal waste was found to have a lower level of specific activity than the limit value. However, Zircaloy cladding contaminated by fission products and actinides during the decladding process of pyroprocessing was revealed to have 52 and 2 times higher specific activity levels than the limit values for alpha and $^{90}Sr$, respectively. Finally, it was found that 88.7% of the metal waste from the 17${\times}$17 Korean Optimized Fuel Assembly design should be disposed of in a deep geological repository. Therefore, it can be summarized that separation technology with a higher decontamination factor for transuranics and strontium should be developed for the efficient management of metal waste resulting from pyroprocessing.

• 한국신재생에너지학회:학술대회논문집
• /
• 2008.05a
• /
• pp.460-463
• /
• 2008

A large part of the overall industrial energy is dissipated as waste heat despite of much development in the utilization of thermal energy. A mean efficiency is reported to be only around 30 to 35%. The existing waste heat recovery technology has reached its limit and consequently, the development of a new technology is necessary. Improving efficiency using thermoelectric technology has recently come into the spotlight because of its unique way to recover thermal energy. In fact, thermoelectric generator directly converts thermal energy into electric energy by a solid state without any moving parts. Futhermore remarkable improvement in the thermoelectric energy conversion efficiency has been achieved. In this study, a thermoelectric generator was made using commercialized thermoelectric modules. With thermoelectric modules attached on a duct surface, hot air was blown into the duct using a hot air blower. On the other side of the module, a water jacket was attached to cool the module. With different air inlet temperatures and water flowrates, the electrical power of the thermoelectric generator was measured.

• Journal of Korean Society of Water and Wastewater
• /
• v.27 no.5
• /
• pp.581-590
• /
• 2013

A research was performed for zero discharge system of waste water which is produced from energy recovery process of waste and biomass. Leachate and all kinds of waste water should be separated and integrated into three categories in addition to converting existing leachate treatment facility into waste water treatment facility as well as introducing a management system of reverse osmosis membrane facility and bioreactor landfill. Following these conditions to better water treatment process, it was likely to produce over 3,000 tons of low-grade recycling water and 2,000 tons of high-grade recycling water per day when zero discharge system of waste water is applied starting from 2016. Economical efficiency was also surveyed in total treatment fee. Present system costs 18,129 million won per year, and suggested zero discharge system would cost 15,789 million won per year.

• Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT)
• /
• v.16 no.1
• /
• pp.93-105
• /
• 2018

In Korea, a huge amount of radioactive concrete waste will be generated through decommissioning of nuclear facilities in the near future; therefore, optimum technology for the treatment of concrete waste should be reviewed thoroughly and the future direction of technology development should be discussed. In this paper, many domestic and foreign examples of generation of radioactive concrete waste were pieced together and the characteristics of radioactive concrete waste were examined. Moreover, we reviewed trends in technology development by analyzing the examples of various studies and practical applications of treatment technologies, such as mechanical decontamination, chemical decontamination, volume reduction, recycling and solidification, and also tried to understand the limitations of existing technologies and determine a direction for technical improvement.

• Nuclear Engineering and Technology
• /
• v.52 no.4
• /
• pp.797-801
• /
• 2020

Most of irradiated graphite that should be disposed comes from moderators and reflectors of nuclear power plants. The quantity of irradiated graphite could be higher in the future if high-temperature reactors (HTRs) will be deployed. In this case noteworthy quantities of fuel pebbles containing semi-graphitic carbonaceous material should be added to the already existing 250,000 tons of irradiated graphite. Industry graphite is largely used in industrial applications for its high thermal and electrical conductivity and thermal and chemical resistance, making it a valuable material. Irradiated graphite constitutes a waste management challenge owing to the presence of long-lived radionuclides, such as ¹⁴C and ³⁶Cl. In the ENEA Nuclear Material Characterization Laboratory it has been successfully designed a procedure based on the exfoliation process organic solvent assisted, with the purpose of investigate the possibility of achieving graphite significantly less toxic that could be recycled for other purpose [1]. The objective of this paper is to evaluate the possibility of the scalability from laboratory to industrial dimensions of the exfoliation process and provide the prototype of a chemical plant for the treatment of irradiated graphite.

• Journal of Korea Society of Waste Management
• /
• v.35 no.7
• /
• pp.587-599
• /
• 2018

The coffee grounds generated during the coffee extraction process contain several resources, but the technology for their recycling has not been commercialized yet, causing various environmental problems. Due to the recent increase in coffee consumption worldwide, the amount of coffee grounds produced has been continuously increasing, reaching more than 750 million tons. In Korea, about 120,000 tons of coffee waste are annually generated; however, most of them are landfilled or incinerated. Although there is still a shortage of coffee waste recycling technologies compared to the amount of coffee grounds produced, various recycling approaches are being actuated in many countries including Korea. In this study, the generation of coffee grounds at home and abroad, the status of coffee grounds recycling, and the associated technology development trends were investigated. The coffee grounds recycling has been studied in the fields of energy, adsorbent, construction, agriculture, and bio-foods. Research is most active in the energy and biotechnology areas; in particular, since the oil in the coffee grounds is valuable as a feedstock for biomass energy, the technology related to energy recovery is currently under development worldwide. Removed because confusing and unnecessary.

• Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT)
• /
• v.20 no.2
• /
• pp.185-192
• /
• 2022

A compacted bentonite buffer is a major component of engineered barrier systems, which are designed for the disposal of high-level radioactive waste. In most countries, the target temperature required to maintain safe functioning is below 100℃. If the target temperature of the compacted bentonite buffer can be increased above 100℃, the disposal area can be dramatically reduced. To increase the target temperature of the buffer, it is necessary to investigate its properties at temperatures above 100℃. Although some studies have investigated thermal-hydraulic properties above 100℃, few have evaluated the water suction of compacted bentonite. This study addresses that knowledge gap by evaluating the water suction variation for compacted Korean bentonite in the 25-150℃ range, with initial saturations of 0 and 0.22 under constant saturation conditions. We found that water suction decreased by 5-20% for a temperature increase of 100-150℃.