작물을 재배하는 데 필요한 여러 가지 환경 조건 중 광은 개 화와 밀접한 연관이 있다. 본 연구는 식용화, 매리골드 화아분 화에 영향을 주는 최적의 광주기를 구명하여 완전제어형 식물 공장에서 효율적으로 재배하기 위해 진행되었다. 실험에 사 용된 광주기는 4, 8, 12, 16시간, 총 4가지로 설정하였다. 매리 골드 ‘듀란고 레드’ 종자를 우레탄 스펀지에 파종한 직후부터 광주기를 처리하였다. 화아분화는 꽃봉오리가 약 2mm 이상 일 때 화아분화가 되었다고 정의하였고, 2－3일 간격으로 조 사하였다. 생육 조사는 지상부의 생체중, 건물중, 초장, 엽면 적을 조사하였다. 최적의 광주기는 식물체의 50%가 화아분 화 된 날을 기준으로 정의하였다. 4시간 처리구에서는 식물체 가 제대로 자라지 못하며 화아도 형성되지 않았다. 8시간 이상 의 처리구에서부터 식물체가 정상적으로 생장하고 화아분화 가 이루어졌지만, 8시간 처리구는 12시간 이상의 처리구에 비 해 화아분화가 더디게 일어났다. 반면에 12시간 처리구와 16 시간 처리구는 서로 유의하지 않은 결과를 보였다. 모든 생육 조사 항목에서 16시간 광주기 처리구가 가장 높은 값을 나타 냈으나 지상부의 건물중과 엽면적을 제외한 나머지 항목에서 12시간 처리구와 유의하지 않았다. 실험 결과에 따르면, 8시간 광주기에서도 화아분화가 일어났지만, 화아형성까지의 시 간이 12시간 이상의 광주기일 때보다 더 많이 소요되었으며, 식물체의 생육 또한 12시간 이상의 광주기를 조사받은 식물 체보다 낮게 나타났다. 본 실험에서 에너지 소비량을 고려한 최적의 매리골드 ‘듀란고 레드’의 광주기는 12시간으로 판단 된다.

PURPOSES : The objective of this study was to review roadway management strategies that can be utilized in the event of a radiological emergency, select feasible alternatives, and simulate a portion of the West Coast network to analyze the effectiveness of these strategies. METHODS : The methodology of the study involved reviewing the relevant literature, extracting the implications, establishing an analysis procedure, and selecting an effectiveness evaluation scale. Using a national transportation database, a network was constructed using Toba, a macroscopic model. RESULTS : A reverse-flow lane system was applied to the West Coast Expressway Glory IC–Hampyeong IC (total 25 km), and a plan to increase the number of lanes was applied to the Seokgyo Street–Shinpyeong Intersection on National Route 23 (total extension 28 km). Consequently, both road management strategies were found to be effective. CONCLUSIONS : This study examined roadway management strategies that can be utilized in the event of a radiological emergency, selected feasible alternatives, and simulated a portion of the West Coast network to verify the effectiveness of these strategies. In the case of reverse flow lanes, it is most effective when applied to expressways that can restrict the entrance and exit of vehicles. In the case of increasing the number of lanes, it is most effective when applied to general roads, and institutional arrangements should be made to enable two-way traffic to use the reserved shoulder lanes.

최근 정원에 대한 관심이 증가함에 따라 자생 식물을 활용 하기 위한 연구가 진행되고 있다. 참식나무(Neolitsea sericea (Blume) Koidz).는 관상용뿐만 아니라 목재로도 이용할 수 있 다. 따라서 이 연구는 참식나무를 정원 식물 소재로 활용하기 위한 대량 증식 기술 개발을 위하여 종자의 발아 특성과 저장 특성에 대해 알아보고자 진행하였다. 2022년 11월 2일 제주 지역에서 채집한 종자를 실험에 사용하였다. 종자의 내·외부 형태 관찰을 한 결과, 종자가 탈리되는 시점에 배가 완전히 발 달한 형태를 가지고 있기 때문에 형태적휴면(morphological dormancy, MD)이 없음을 확인할 수 있었다. 수분흡수 실험 을 통하여 참식나무 종자는 72시간 이후 40% 정도의 수분흡수 율을 보여 내과피의 불투성에 의한 물리적휴면(PY, physical dormancy)이 없는 것으로 판단하였다. 채종 후 곧바로 무가 온 온실에 파종하면 이듬해 5월에 약 60%가 발아하였다. 저 온습윤처리 0, 2, 4, 8, 12주 후 25/15°C에 배양한 결과 최 종 발아율은 각각 0%, 0%, 21.1%, 32.6%, 34.7%였다. GA3 를 0, 10, 100, 1000mg·L-1 농도로 처리한 후 25/15°C에서 배양한 결과, 최종 발아율은 각각 0, 14.5, 22.7, 42.2%였다. 따라서 본 실험의 결과 8주 이상의 저온 처리 또는 GA3 처리 를 통하여 어느 정도 휴면을 타파시킬 수 있었다. 이러한 결과 를 통하여 종자는 intermediate PD를 가지는 것으로 판단 된다. 일부 발아하지 않은 종자는 deep PD를 가지고 있는 것 으로 판단하였다. 후숙 처리 실험의 결과 4주 만에 수분 함량 이 2% 이하로 감소하였고 종자는 전혀 발아하지 않았다. 따라 서 종자의 저장 특성 중 recalcittrant 특성을 지닌 난저장성 종자로 보인다.

Stomatal pore is an important physiological trait that is closely linked to photosynthesis and transpiration as carbon dioxide and water vapor move through it between the atmosphere and plants. The present study investigated stomatal traits, such as stomatal density, index and size, of herbaceous native and alien plant species living in a riparian park on the Nakdong River to understand how those traits vary and to know if successful settlement of alien plants is attributed to those traits. There was no difference in stomatal density, index and size between native and alien plants with kidney-shaped stomata, suggesting that an empty ecological niche is not an essential prerequisite for the successful settlement of alien plants. Stomatal density showed a negative correlation with leaf thickness and leaf dry weight content (LMDC), but there was no correlation with Specific leaf area (SLA). All plants with kidney-shaped stomata had amphistomatous leaves, and the density and size of dumbell-shaped stomata were lower than those of kidney-shaped stomata.

The limitation in deriving the species richness representing the entire country of South Korea lies in its relatively short history of species field observations and the scattered observation data, which has been collected by various organizations in different fields. In this study, a comprehensive compilation of the observation data for plants held by agencies under the Ministry of Environment was conducted, enabling the construction of a time series dataset spanning over 100 years. The data integration was carried out using minimal criteria such as species name, observed location, and time (year) followed by data verification and correction processes. Based on the integrated plant species data, the comprehensive collection of plant species in South Korea has occurred predominantly since 2000, and the number of plant species explored through these surveys appears to be converging recently. The collection of species survey data necessary for deriving national-level biodiversity information has recently begun to meet the necessary conditions. Applying the Chao 2 method, the species richness of indigenous plants estimated at 3,182.6 for the 70-year period since 1951. A minimum cumulative period of 7 years is required for this estimation. This plant species richness from this study can be a baseline to study future changes in species richness in South Korea. Moreover, the integrated data with the estimation method for species richness used in this study appears to be applicable to derive regional biodiversity indices such as for local government units as well.

2015년부터 2022년도까지 6개목(딱정벌레목, 노린재목, 나비목, 벌목, 파리목, 총채벌레목) 곤충들에 대해서 식물검역현장 검출실적과 국내 보고된 미기록종을 분석하였다. 해당기간 동안 국경검역에서 6개목 곤충은 총 45,084건이 검출되었다. 같은 기간 국내에서는 총 545종이 미기록종 으로 보고되었으며, 이중 9종은 국경검역에서도 검출된 것으로 확인되었다. 검역현장에서는 딱정벌레목, 총채벌레목, 노린재목이 높은 검출률을 보 였으며, 국내 미기록종 중에서는 벌목이 176종으로 가장 많이 보고되었다. 본 연구를 통해 침입압력(국경검역 검출)과 실제 침입(국내 미기록종 발 견) 사이에 비동시성이 확인되었다. 향후 보다 장기적인 분석이 필요할 뿐만 아니라 지속적인 식물검역시스템 개선이 필요할 것으로 판단된다.

To construct and operate nuclear power plants (NPPs), it is mandatory to submit a radiation environmental impact assessment report in accordance with Article 10 and Article 20 of the Nuclear Safety Act. Additionally, in compliance with Article 136 of the Enforcement Regulations of the same law, KHNP (Korea Hydro & Nuclear Power) annually assesses radiation environmental effects and publishes the results for operating NPPs. Furthermore, since the legalization of emission plans submission in 2015, KHNP has been submitting emission plans for individual NPPs, starting with the Shin-Hanul 1 and 2 units in 2018. These emission plans specify the emission quantities that meet the dose criteria specified by the Nuclear Safety and Security Commission. Before 2002, KHNP used programs developed in the United States, such as GASPAR and LADTAP, for nearby radiation environmental impact assessments. Since then, KHNP has been using K-DOSE60, developed internally. K-DOSE60 incorporates environmental transport analysis models in line with U.S. regulatory guidance Regulatory Guide 1.109 and dose assessment models reflecting ICRP-60 recommendations. K-DOSE60 is a stand-alone program installed on individual user PCs, making it difficult to manage comprehensively when program revisions are needed. Additionally, during the preparation of emission plans and the licensing phase, improvements to KDOSE60’ s dose assessment methodology were identified. Furthermore, in 2022, regulatory guidelines regarding resident dose assessments were revised, leading to additional improvement requirements. Currently, E-DOSE60, being developed by KHNP, is a network-based program allowing for integrated configuration management within the KHNP network. E-DOSE60 is expected to be developed while incorporating the identified improvements from K-DOSE60, in response to emission plan licensing and regulatory guideline revisions. Key improvements include revisions to dose assessment methodologies for H-13 and C-14 following IAEA TRS-472, expansion of dose assessment points, and changes in socio-environmental factors. Furthermore, data such as site meteorological information and releases of radioactive substances in liquid and gaseous forms can be linked through a network, reducing the potential for human errors caused by manual data entry. Ultimately, E-DOSE60 is expected to optimize resident exposure dose assessment and enhance public trust in NPP operation.

In the dismantling of nuclear power plants, various forms of radioactive gaseous waste are generated when cutting concrete and metal structures. Large amounts of radioactive dust and aerosols generated during the cutting process of each structure can cause radiation exposure to the environment around the workplace and to the radiation exposure in the body of workers. When cutting structures, water is sprayed to reduce the generation of aerosols, so early saturation of the filter is expected due to radioactive aerosols and fine particles containing a large amount of moisture. A mobile air purification device is being developed to a fast and efficient air purifier that can be used for a long time operation to protect workers from radiation exposure in high radiation areas and to minimize the amount of secondary waste generated. In this paper, the direction for a new concept of unit technology that can achieve the development purpose is described.

Thermal cutting processes that can be applied to dismantling nuclear power plants include oxygen cutting, plasma cutting, and laser cutting. According to the global trend, research projects are being carried out in various countries to upgrade laser cutting, and many studies are also being conducted in Korea with plans to apply laser cutting processes when dismantling nuclear power plants. However, with the current technology level of the laser cutting process, the maximum thickness that can be cut is limited to 250 mm. Therefore, in this study, a laser-oxygen hybrid cutting process was implemented by adding a laser heat source to the oxygen cutting process that can cut carbon steel with a thickness of 250 mm or more (RV, beam, column, beam, etc.) when dismantling the nuclear power plant. This has the advantage of improving the cutting speed and reducing the cutting width Kerf compared to conventional oxygen cutting. In this research, the laser-oxygen hybrid cutting process consisted of laser cutting to which Raycus’ 8 kW Fiber Laser power source was applied and oxygen cutting to which hydrogen was applied with Fuel Gas. The oxygen torch was placed perpendicular to the test piece, and the laser head was irradiated by tilting 35° to 70°. The effects of cutting directions on quality and performance were studied, and cutting paths were selected by comparing cutting results. Thereafter, it was confirmed that there is an optimal laser output power according to the cutting thickness by studying the effect on the cutting surface quality by changing only the laser output power under the same cutting conditions. The results of this study are expected to be helpful in the remote cutting process using laser-oxygen hybrid cutting when dismantling domestic nuclear power plants in the future.

When dismantling a power plant, a large amount of radioactive tanks are generated, and it is estimated that a significant amount of sludge will accumulate inside the tanks during long-term operation. In the process of dismantling a radioactive tanks, it is important to know the composition of the sludge because the sludge present inside must first be removed and then disposed of. In the case of certain tanks, it can be predicted that corrosion products generated due to system corrosion are the main cause of sludge formation. However, in the case of some tanks, it is not easy to predict the sludge composition because various dispersed particles in addition to corrosion products may be mixed with the wastewater. Even if it is collected and analyzed, the sludge composition can change significantly depending on the operation history, so the analysis results cannot be considered representative of the composition. In the case of LHST, surfactant components introduced during the washing and shower process, oil components and dispersed particles dissolved by the surfactant accumulate inside the tank, making sludge difficult to remove. In addition, even if it is removed by ultra-high pressure spraying, unexpected problems may occur in the subsequent treatment process due to the surfactant contained therein. Therefore, it is necessary to analyze in more detail the characteristics of sludge accumulated in LHST and prepare countermeasures. A test procedure was prepared to evaluate the characteristics of sludge accumulating in LHST. According to the test results, the long-term sludge accumulation tendency of the LHST is summarized as follows. ① Initially, the sludge settling speed increases slowly until a surface sludge layer is formed. ② After the surface sludge layer is formed, the sludge rapidly settles until the sludge layer becomes somewhat thicker. ③ When the sludge layer is formed to a certain extent, the sludge escape rate increases and the sludge accumulation rate decreases again. It is assumed that the sludge escape speed is closely related to the fluid flow speed in the relevant area. It is believed that the combined effect of these phenomena will determine the thickness of the sludge layer that will accumulate inside the tank, but it was not possible to evaluate how much the sludge layer would accumulate based on the experimental results alone. However, it can be assumed that significant sludge accumulation occurred in areas where fluid flow was minimal and sludge formation nuclei easily accumulates.

Kori Unit 1 nuclear power plant is a pressurized water reactor type with an output of 587 Mwe, which was permanently shut down on June 18, 2017. Currently, the final decommissioning plan (FDP) has been submitted and review is in progress. Once the FDP is approved, it is expected that dismantling will begin with the secondary system, and dismantling work on the primary system of Kori Unit 1 will begin after the spent nuclear fuel is taken out. It is expected that the space where the secondary system has been dismantled can be used as a temporary storage place, and the entire dismantling schedule is expected to proceed without delay. The main equipment of the secondary system is large and heavy. The rotating parts is connected to a single axis with a length of about 40 meters, and is complexly installed over three floors, making accessibility very difficult. A large pipe several kilometers long that supplies various fluids to the secondary system is installed hanging from the ceiling using a hanger between the main devices, and the outer diameter of the pipe is wrapped with insulation material to keep warm. In nuclear secondary system decommissioning, it is very important to check for radiation contamination, establish and implement countermeasures, and predict and manage safety and environmental risks that may occur when cutting and dismantling large heavy objects. So we plan to evaluate the radiation contamination characteristics of the secondary system using ISOCS (In- Situ Object Counting System) to check for possible radioactive contamination. According to the characteristics results, decommissioning plans and methods for safe dismantling by workers were studied. In addition, we conducted research on how to safely dismantle the secondary system in terms of industrial safety, such as asbestos, cutting and handling of heavy materials and so on. This study proposes a safe decommissioning method for various risks that may occur when dismantling the secondary system of Kori Unit 1 nuclear power plant.

After the major radioactivation structures (RPV, Core, SG, etc.) due to neutron irradiation from the nuclear fuel in the reactor are permanently shut down, numerous nuclides that emit alpha-rays, beta-rays, gamma-rays, etc. exist within the radioactive structures. In this study, nuclides were selected to evaluate the source term for worker exposure management (external exposure) at the time of decommissioning. The selection of nuclides was derived by sequentially considering the four steps. In the first stage, the classification of isotopes of major nuclides generated from the radiation of fission products, neutron-radiated products, coolant-induced corrosion products, and other impurities was considered as a step to select evaluation nuclides in major primary system structures. As a second step, in order to select the major radionuclides to be considered at the time of decommissioning, it is necessary to select the nuclides considering their half-life. Considering this, nuclides that were less than 5 years after permanent suspension were excluded. As a third step, since the purpose of reducing worker exposure during decommissioning is significant, nuclides that emit gamma rays when decaying were selected. As a final step, it is a material made by radiation from the fuel rod of the reactor and is often a fission product found in the event of a Severe accident at a nuclear power plant, and is excluded from the nuclide for evaluation at the time of decommissioning is excluded. The final selected Co-60 is a nuclide that emits high-energy gamma rays and was classified as a major nuclide that affects the reduction of radiation exposure to decommissioning workers. In the future, based on the nuclide selection results derived from this study, we plan to study the evaluation of worker radiation exposure from crud to decommissioning workers by deriving evaluation results of crud and radioactive source terms within the reactor core.

In nuclear power plant (NPP) decommissioning, ventilation and purification of the building atmosphere are important to create a working environment, ensure worker safety, and prevent the release of gaseous radioactive materials into the environment. The heating, ventilation, and air conditioning (HVAC) system of each building is maintained, modified, or newly installed. In this study, based on APR1400, operation strategies were presented in case of ventilation abnormalities in the reactor containment building (RCB), where highly radioactive particles and high dust are most frequently generated during NPP decommissioning. For research, it was assumed that the entire RCB atmospheric ventilation during decommissioning would use the RCB purge system of the existing NPP and perform continuous ventilation. Additionally, it is assumed that areas where high radiation particles and high dust occur locally, such as reactor containers or internal segments, are sealed with tents and purified using a HEFA filter of a temporary portable HVAC, and a exhaust flow path is connected to the discharge duct of the existing RCB purge system. The possibility of abnormal occurrence was largely divided into two cases. First, when large amounts of uncontrolled pollutants are released into the atmosphere inside the RCB, discharge to the environment is stopped manually or automatically by a modified engineered safety function activation signal (ESFAS). Afterwards, the RCB purge system should be operated in recirculation mode to sufficiently purify the RCB atmosphere with a HEPA filter. Second, when the first train of the low volume purge system is not running due to a failure, standby train should be operated. If both low volume purge trains fail, a high volume purge system is used. Intermittent purge operation is preferred due to large capacity during high volume purge operation. In cases where it is not possible to operate all purge systems due to common issues such as power supply, atmospheric sampling is performed to determine whether to proceed with the work inside RCB.

The radiological characterization of SSCs (Structure, Systems and Components) plays one of the most important role for the decommissioning of KORI Unit-1 during the preparation periods. Generally, a regulatory body and laws relating to the decommissioning focus on the separation and appropriate disposal or storage of radiological waste including ILW (intermediate level waste), LLW (low level waste), VLLW (very low level waste) and CW (clearance waste), aligned with their contamination characteristics. The result of the preliminary radiological characterization of KORI Unit-1 indicated that, apart from neutron activated the RV (reactor vessel), RVI (reactor vessel internals), and BS (biological shielding concrete), the majorities of contamination were sorted to be less than LLW. Radiological contamination can be evaluated into two methods. Due to the difficulties of directly measuring contamination on the interior surfaces of the pipe, called CRUD, the assessment was implemented by modeling method, that is measuring contamination on the exterior surfaces of the pipes and calculating relative factors such as thickness and size. This indirect method may be affected by the surrounding radiation distribution, and only a few gamma nuclides can be measured. Therefore, it has limitation in terms of providing detailed nuclide information. Especially, α and β nuclides can only be estimated roughly by scaling factors, comparing their relative ratios with the existing gamma results. To overcome the limitation of indirect measurement, a destructive sampling method has been employed to assess the contamination of the systems and component. Samples are physically taken some parts of the systems or components and subsequently analyzed in the laboratory to evaluate detailed nuclides and total contamination. For the characterization of KORI Unit-1, we conducted the radiation measurement on the exterior surfaces of components using portable instruments (Eberline E-600 SPA3, Thermo G20-10, Thermo G10, Thermo FH40TG) at BR (boron recycle system) and SP (containment spray system) in primary system. Based on these results, the ProUCL program was employed to determine the destructive sample collection quantities based on statistical approach. The total of 5 and 8 destructive sample quantities were decided by program and successfully collected from the BR and SP systems, respectively. Samples were moved to laboratory and analyzed for the detail nuclide characteristics. The outcomes of this study are expected to serve as valuable information for estimating the types and quantities of radiological waste generated by decommissioning of KORI Unit-1.

The purpose of this report is to provide a summary of the Phase 1 Final Status Survey (FSS) Final Report results and overall conclusions which conduct that the Zion Nuclear Power Station (ZNPS) facility and site meets the 25 mrem(0.25 mSv)per year release criterion as established in Nuclear Regulatory Commission Regulation (NRC) 10 CFR 20.1402 “Radiological Criteria for Unrestricted Use”. The FSS results provided assessment and summarize that any residual radioactivity results in a Total Effective Dose Equivalent (TEDE) to an Average Member of the Critical Group (AMCG) that does not exceed 25 mrem per year, and the residual radioactivity has been reduced to levels that are as low as reasonably achievable (ALARA). The release criterion is translated into site-specific Derived Concentration Guideline Levels (DCGLs) for assessment and summary. ZionSolutions, a decommissioning service provider, estimates that a total of four (4) FSS Final Reports be generated and submitted to the NRC during the decommissioning project. ZionSolutions established the Characterization/License Termination (C/LT) Group, within the Radiation Protection division, with sufficient management and technical resources to fulfill project objectives. The C/LT Group is responsible for the safe completion of all surveys related to characterization and final site closure. Approved site procedures and detailed Technical Support Documents (TSD) direct the FSS process to ensure consistent implementation and adherence to applicable requirements. The development and planning phase was initiated in 1999 by the “Zion Station Historical Site Assessment” (HSA) and the initiation of the characterization process for FSS. Develop the information necessary to support FSS design, including the development of Data Quality Objectives (DQOs) and survey instrument performance standards. DQOs are qualitative and quantitative statements derived from the DQOs process that clarify technical and quality objectives. The next step, FSS design utilizes the combination of traditional scanning surveys, systematic sampling protocols and investigative/judgmental methodologies to evaluate survey units relative to the applicable release criteria for open land sample plans. To aid in the development of an initial suite of potential radionuclides of concern for the decommissioning of ZNPS, the analytical results of representative characterization samples collected at the site were reviewed. At this FSS design step, the Radionuclides of Concern (ROC) are determined. As Co-60 and Cs-137 account for 99.5% of the analysis results of concrete core sampling data form ZNPS’s Containment Building and Auxiliary Building, they are determined and used as the basic ROC in the survey design. Additionally, site information is described and Historical Site Assessment (HSA) is performed. Data collected for the initial HSA will be used to establish the initial regional survey unit and corresponding MARSSIM classification. Next, an assessment of the collected data is performed using the DQO process, and a survey methodology is established by selecting a sampling method and measuring instrumentation. These result judgments provide guidance for C/LT Engineer to interpret findings using the Data Quality Assessment (DQA) process, which analysis Recorded data, Missing values, Deviation from established procedure, and Analysis flags. In conclusion, FSS is the process used to demonstrate that the ZNPS facility and site comply the radiological criteria for unrestricted use specified in 10 CFR.20. The purpose of FSS Sample Plan is to describe the methods to be used in planning, designing, conducting, and evaluating the FSS.