왕담배나방은 기주범위가 넓은 다식성으로 토마토, 옥수수, 담배, 해바라기, 땅콩, 면화, 콩 등 다양한 작물에 피해를 주는 해충이며, 장거리 이동성 해충으로도 유명하다. 본 연구에서는 기존 자료를 바탕으로 월동번데기의 성충으로 우화모델(2종)를 작성하고 포장 실측자료와 비교하여 평가하였다. 월동 번데기는 휴면을 종료한 후 안점소실 과정을 거쳐 정상의 번데기로 이행되고 비로소 성충으로 우화하는 것으로 가정하였다. 첫 번째 모형(모 형 1)은 2단계 휴면후발육 우화모형으로 안점소실에 필요한 적산온도 59.9DD (발육영점온도 15.9℃)와 번데기 발육완료에 필요한 192.3DD (발육영점온도 10.2℃)로 구성하였다. 두 번째 모형(모형 2)은 온도자극 우화모형으 로 17.63℃의 온도자극 이후에 번데기 발육을 개시하는 모형이었다. 적산온도는 단순 평균온도법과 사인곡선법 으로 계산하여 비교하였다. 포장자료와 비교 결과 모형 1에 사인곡선법을 적용했을 때 실측치와 편차가 3일보다 작거나 차이가 없었다. 반면 사인곡선법을 적용한 모형 2는 3~6.5일의 편차가 있었다. 평균온도법을 적용한 경우 두 모형 모두 실측치와 편차가 증가하였다. 모형을 활용하여 예측 우화시기와 트랩유살자료를 비교하여 왕담배 나방의 비래 가능성에 대하여 고찰하였다.

The genus Kurahashiodes is a monotypic genus that only recorded in Japan with Kurahashiodes suenagai. This species was first known by Wohlfahrtiodes suenagai in 1994, but reorganized with the new genus Kurahashiodes by Verves in 2001. Kurahashiodes was close to genus Wohlfahrtia but can be distinguished by R1 haired, surstylus narrow, phallus without ventral protuberance. This genus is reported for the first time in Korean fauna. The diagnosis and photographs of the Kurahashiodes suenagai are provided herein.

표면발현(surface-display system)은 세포 또는 바이러스 표면에 목적 단백질을 고정하여 발현시킴으로써 목적 단백질에 대하여 독립적인 공간 구조 및 생물학적 활성을 부여하는 단백질 공학 기술이다. 또한 이를 이용하여 높은 중화항체 유도 및 대량생산이 가능한 삼량체의 형태로 항원 단백질의 발현 또한 가능하다. BES(baculovirus expression system)에서의 표면발현 기술은 번역 후 수정과정 및 복잡한 구조의 다양한 단백질의 발현이 가능하기 에 다른 숙주 기반 시스템보다 효율적이라고 보고되고 있다. 그러나 목적 단백질 외의 다른 표면 단백질과 발현 공간에서의 경쟁으로 목적 단백질의 낮은 생산량이 큰 문제점으로 지적되고 있다. 따라서, 이러한 BES에서 표면 발현의 생산 효율을 증대시키기 위하여, 동일한 표면 공간에 대한 단백질 간의 발현 경쟁에 대해 실험적으로 확인 후, 그를 해결하기 위하여 표면발현에 최적인 목적 단백질 발현을 위한 프로모터 선발 실험을 수행하였다. 이를 통해 BES에서 표면발현에 의한 목적 단백질의 생산 효율을 증대시킬 수 있음을 확인하였다.

Corn silage is extensively utilized in ruminant feeding on a global scale, with substantial research efforts directed towards enhancing its nutritional worth and managing moisture content. The purpose of this study was to assess the impact of normal cutting height and elevated cutting height on whole-crop corn silage. Corn was harvested at heights of 15 cm and 45 cm above the ground, respectively, 45 days after heading. The harvested corn was cut into 2-3 cm lengths and packed into 20-liter plastic silos in triplicate. The results showed that dry matter (DM), crude protein (CP), water soluble carbohydrates (WSC), and in vitro dry matter digestibility (IVDMD) of C45 were significantly higher than those of the control, while the neutral detergent fiber (NDF) was significantly lower in C45 (p<0.05). The C15 had higher yields than C45 (p<0.05). There was no significant difference in the total digestible nutrients (TDN) yield of whole-crop corn silage. The increase in cutting height resulted in a larger change in moisture content and NDF per centimeter. After 60 days-ensiling, C45 showed significantly lower NH3-N concentrations. Moreover, C45 had significantly higher lactic acid concentration, lactic acid/acetic acid ratio, and lactic acid bacteria count compared to the control. Mold was not detected and the yeast count was less than 2 log10 cfu/g fresh matter in both control and C45. In summary, C45 improved the feeding value and fermentation quality of whole-crop corn silage at the expense of forage productivity.

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.

The decommissioning of Korea Research Reactor Units 1 and 2 (KRR-1&2), the first research reactors in South Korea, began in 1997. Approximately 5,000 tons of waste will be generated when the contaminated buildings are demolished. Various types of radioactive waste are generated in large quantities during the operation and decommissioning of nuclear facilities, and in order to dispose of them in a disposal facility, it is necessary to physico-chemically characterize the radioactive waste. The need to transparently and clearly conduct and manage radioactive waste characterization methods and results in accordance with relevant laws, regulations, acceptance standards is emerging. For radioactive waste characterization information, all information must be provided to the disposal facility by measuring and testing the physical, chemical, and radiological characteristics and inputting related documents. At this time, field workers have the inconvenience of performing computerized work after manually inputting radioactive waste characterization information, and there is always a possibility that human errors may occur during manual input. Furthermore, when disposing of radioactive waste, the production of the documents necessary for disposal is also done manually, resulting in the aforementioned human error and very low production efficiency of numerous documents. In addition, as quality control is applied to the entire process from generation to treatment and disposal of radioactive waste, it is necessary to physically protect data and investigate data quality in order to manage the history information of radioactive waste produced in computerized work. In this study, we develop a system that can directly compute the radioactive waste characterization information at the field site where the test and measurement are performed, protect the stored radioactive waste characterization data, and provide a system that can secure reliability.

The domestic representative nuclear fuel cycle facilities are post-irradiation examination facility (PIEF) and Irradiated Examination Facility (IMEF) at KAERI. They have regularly operated since 1991 and 1993, respectively. Due to the long period of use, the facilities are ageing, and maintenance costs are increasing every year. The maintenance methods have mainly been breakdown maintenance (BM) and partially preventive maintenance (PM). They involve replacing components that have problems through periodic inspections by on-site inspectors. However, these methods are not only uncertain in terms of replacement cycles due to worker’s deviation on the inspection results, but also make it difficult to respond accidents developed through failures on the critical equipment that confines radioactive material. Therefore, an advanced operation and maintenance studied in 2022 through all of nuclear facilities operated at KAERI. Advancement strategy in four categories (safety, sustainability, performance, innovativeness) was analyzed and their priorities according to a facility environment were determined so a roadmap for advanced operation and maintenance could be developed. The safety and sustainability are higher importance than the performance and innovativeness because facilities at KAERI has an emphasis on research and development rather than industrial production. Thus, strategy for advancement has focused even more on strengthening the safety and sustainability. To enhance safety, it has been identified that immediate improvement of aged structures, systems, and components (SSCs) through large-scale replacement is necessary, while consideration of implementing an ageing management program (AMP) in the medium to long term is also required. Facility sustainability requires strengthening operation expertise through training, education, and cultivation of specialized personnel for each system, and addressing outstanding regulatory issues such as approval of radiation environment report on the nuclear fuel processing facilities and improvement work according to fire hazard analysis. One of the safety enhancement methods, AMP, is a new maintenance approach that has not been previously applied, so it had to be thoroughly examined. In this study, an analysis was conducted on the procedure and method for introducing an AMP. An AMP for nuclear fuel cycle facilities was developed by analyzing the AMP applied to the BR2 research reactor in Belgium and modifying it for application to nuclear fuel cycle facilities. The ageing management for BR2 has the objective to maintain safety, availability and cost efficiency and three-step process. The first step is the classification of SSCs into four classes to apply graded approach. Secondly, ageing risk is assessed to identify critical failure modes, their frequency and precursors. Final step involves defining measures to reduce the ageing risk to an acceptable level in order to integrate the physical and economic aspects of ageing into a strategy for inspection, repair, and replacement. Similar approach was applied to the nuclear fuel cycle facility. Firstly, the SSCs of nuclear fuel cycle facilities have been classified according to their safety and quality classifications, as well as whether they are part of the confinement boundary. The SSCs involved in the confinement boundary were given more weight in the classification process, even if they are not classified as safety-class. A risk index for ageing was introduced to determine which prevention and mitigation measure should be chosen. By multiplying the health index and the impact index, the ageing risk matrix provides a numerical score that represents guidance on the prevention and mitigation of ageing effect. The health index is determined by combining the likelihood of failure and engineering evaluation of the current condition of SSCs, whereas the impact index is calculated by taking into account the severity of consequences and the duration of downtime resulting from a failure. This ageing management has to be thoroughly reviewed and modified to suit each facility before being applied to nuclear fuel cycle facilities.