Korea Atomic Energy Research Institute (KAERI) has been operating the Post Irradiation Examination Facility (PIEF) for fuel examinations. The facility has pools and hot cells for handling and examining fuel assemblies and rods. Among the hot cells, the second cell is for measuring rod internal pressure (RIP) and then cutting the rod to make samples for destructive tests. Currently, the cutting machine is broken, so it has to be replaced. Because the existing cutting machine consists of many parts and its size was quite a bit large to handle and treat for the radioactive waste disposal, the disassembly work has been performed to make it smaller using manipulators. The drawings of the cutting machine were reviewed and the disassembly tools were developed considering workability when the work performed at the hot cell using the manipulators. The large parts such as motor, mirror and cable, etc., were able to be disassembled and the machine size became so smaller that it could be easily handled for the disposal.

Korea Atomic Energy Research Institute (KAERI) has been operating the Post Irradiation Examination Facility (PIEF). The facility has many PIE equipment and one of them is a hydrogen analyzer for measuring hydrogen contents in Zr cladding of spent fuel. The cladding tube of fuel is oxidized in the core environment of high temperature and pressure and absorbs some of the hydrogen generated during the oxidation. The hydrogen content increases with the increase of burn-up, and causes hydriding of the material, which degrades the mechanical properties. Therefore, hydrogen content analysis of the cladding tube is required for the performance and integrity evaluation of spent fuel. In PIEF, the hydrogen analyzer extracts hydrogen gas from Zr cladding by the hot extraction method. The hydrogen gas flows with inert gas and oxidizes to H2O through a CuO reagent. Finally, an IR detector measures the hydrogen amount from the absorbed IR intensity at a specific wavelength. Because the equipment is in the glove box and has some consumable parts, the maintenance work was performed as a radiation work.

본 연구는 주변 환경의 차이에 따른 화분매개곤충의 유입 특성을 파악하기 위하여 국립수목원 내 진화속을걷 는정원과 부추속전문전시원에 식재된 울릉산마늘의 화분매개곤충을 조사하였다. 2023년 5월 22일부터 6월 2일 까지 꽃이 70% 이상 개화하였을 때 포충망을 활용하여 8일간 곤충을 채집하였고, 각 전시원 별 식생(피도), 기후 (온도·습도·조도)를 조사하였다. 조사 결과 진화속을걷는정원에서 피도 60% 온도 26.4℃, 습도 31.5%, 조도 40953.6lx, 화분매개곤충 20과 450개체, 부추속전문전시원은 피도 90%, 온도 25.6℃, 습도 31.6%, 조도 6387lx, 화분매개곤충 15과 196개체로 나타났다. 온도와 조도가 상대적으로 높은 진화속을걷는정원이 채집된 곤충의 다양성과 방문 빈도가 높았다. 시간대별 곤충의 방문 빈도를 비교해본 결과 온도와 조도는 개체수가 증가할 때 같이 증가하는 경향을 보였으며, 습도는 반대의 경향을 보였다.

Golden apple snails(GAS) are native to warm regions such as Central and South America and Southeast Asia, and were first introduced as a high-protein food. GAS are omnivorous and have a habit of eating plants submerged in water, so they have been used for eco-friendly weed control in rice fields since 1992. When the GAS was first introduced, it was thought that it would be impossible to overwinter in Korea. However after 2000, overwintering individuals were founded and damage to rice occurred and the development of means to control GAS has been required. In this study, we tested the effectiveness of an eco-friendly pest control agent using Styrax japonicus that grow naturally in Korea. As a result of exposing GAS to S. japonicus fruit powder, a 100% molluscicidal effect was confirmed at 66.7ppm. To investigate the duration of effect, treatment was performed at the same concentration and molluscicidal effect of more than 90% was up to 3 days after treatment. The killing effect of each part of the S. japonicus was compared, and the seed extracts showed no killing effect at all concentrations, while the sarcocarp extracts showed a 100% killing effect up to 33.3ppm, and the fruit extracts showed a 100% killing effect up to 200ppm.

Natural uranium-contaminated soil in Korea Atomic Energy Research Institute (KAERI) was generated by decommissioning of the natural uranium conversion facility in 2010. Some of the contaminated soil was expected to be clearance level, however the disposal cost burden is increasing because it is not classified in advance. In this study, pre-classification method is presented according to the ratio of naturally occurring radioactive material (NORM) and contaminated uranium in the soil. To verify the validity of the method, the verification of the uranium radioactivity concentration estimation method through γ-ray analysis results corrected by self-absorption using MCNP6.2, and the validity of the pre-classification method according to the net peak area ratio were evaluated. Estimating concentration for 238U and 235U with γ-ray analysis using HPGe (GC3018) and MCNP6.2 was verified by 􀟙-spectrometry. The analysis results of different methods were within the deviation range. Clearance screening factors (CSFs) were derived through MCNP6.2, and net peak area ratio were calculated at 295.21 keV, 351.92 keV(214Pb), 609.31 keV, 1120.28 keV, 1764.49 keV(214Bi) of to the 92.59 keV. CSFs for contaminated soil and natural soil were compared with U/Pb ratio. CSFs and radioactivity concentrations were measured, and the deviation from the 60 minute measurement results was compared in natural soil. Pre-classification is possible using by CSFs measured for more than 5 minutes to the average concentration of 214Pb or 214Bi in contaminated soil. In this study, the pre-classification method of clearance determination in contaminated soil was evaluated, and it was relatively accurate in a shorter measurement time than the method using the concentrations. This method is expected to be used as a simple pre-classification method through additional research.

IAEA safety standards document and international programs (such as BIOMASS) related to the assessment of the biosphere around High Level Radioactive Waste (including Spent Nuclear Fuel) repositories require the assessment of the biosphere to use the assumption that the current natural environment and human society will be maintained, and at the same time, the evolution of the distant future changes also need to be taken into account. In Korea, which has not designated candidate disposal sites, it is necessary to investigate and predict the current state and future changes of the natural environment throughout Korea and apply it practically to Biosphere assessment (for BDCF derivation) for candidate disposal sites suitability assessment and Safety Case (for performance assessment) preparation for design, construction, operation, and post-closure management. To this end, the natural environment in the fields of Topography, Geology, Soil, Ecology, Weather and Climate, Animals and Plants, Hydrology, Ocean, Land-use, etc. and human society in the fields of Population Distribution, Spatial-Planning, Urban Form, Industrial-Structure, Lifestyle etc. are being investigated in the context of current status, past change records, and future change potential in the Korean Peninsula. This paper summarizes those investigations to date. This study referred Biomass-6 [IAEA] and National Atlas I (2019)/II (2020)/III (2021) [National Geographic Information Institute of the Korea Ministry of Land, Infrastructure and Transport].

Korea Atomic Energy Research Institute (KAERI) has been operating the Post Irradiation Examination Facility (PIEF) for spent fuel. The facility has pools and hot cells for handling and examining fuel assemblies and rods. In the first hot cell, non-destructive tests such as visual inspection, defect detection, oxide layer thickness measurement, and gamma scanning are performed on a full-length fuel rod. Then, the fuel rod is transported to the next hot cell for measuring the rod internal pressure (RIP). After the RIP measurement, the fuel rod is cut by a cutting machine to make samples for destructive tests. Currently, the existing cutting machine is broken, so a new machine needed to be designed and manufactured. The major considerations for designing the cutting machine were convenience of remote handling and decontamination. The machine was modularized and its handling parts were designed to be easily controlled by manipulators. The cover was designed to prevent radioactive contamination of the surrounding area.

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.

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.

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.

대부분의 모바일 AR 컨텐츠들은 모바일 디바이스의 기술적 한계로 인해 평면 탐지 후, 그 위에서만 구현되는 제한된 구조를 가지고 있다. 이러한 문제는 제한된 공간이 표현의 범위를 제한하기 때문에 모바일 AR의 확산 에 크게 저해가 될 수 있다. 한편 Unity의 AR Foundation이 제공하는 ‘Meshing’은 실제 오브젝트의 크기와 위 치에 알맞게 메시를 생성해주는데, 이를 활용한다면 모바일 AR 컨텐츠들은 평면에서 벗어나 더 넓은 현실 공 간에 구현될 수 있다. 하지만 ‘Meshing’은 모바일 기기의 센서가 닿지 못하는 부분에는 메시를 생성하지 않기 때문에 별도의 작업 없이 그대로 사용한다면 게임 오브젝트가 빠져나갈 수 있는 구멍이 생길 수 있다. 이 구 멍은 컨텐츠 구현에 있어서 치명적이기에 Hole-Filling 알고리즘을 사용하여 구멍을 메우고자 하는 연구가 있 었다. 하지만 기존 연구에서 사용하는 Hole-Finding 알고리즘은 특정 상황에서 외곽선과 구멍을 제대로 구별해 내지 못하는 문제가 있다. 이 문제는 일부 구멍은 메우지 못하고 외관선끼리 이어버려 컨텐츠에 치명적인 문 제를 야기한다. 본 논문에서는 Meshing이 제공하는 노말 벡터와 경계선들로 계산한 노말 벡터 간의 차이를 이 용해 구멍과 외곽선을 구분하는 방법을 제안한다. 이 방법을 적용한 결과, 이전 연구의 방법보다 좀 더 빠르 면서 구멍과 외곽선을 제대로 구별하는 모습을 확인하였다.

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.