When occurring at a nuclear power plant (NPP) by accidents, accurate prediction and identification of the process of radioactive material dispersing into atmosphere is important to protect public and environment. Atmosphere dispersion of radioactive materials is significantly influenced by wind direction and wind speed. The government and nuclear operator continuously monitor wind data at nuclear sites through meteorological tower to prepare for such accidents involving the release of radioactive materials. The purpose of this study is to construct wind rose diagrams at 5 NPP sites (Kori, Saewool, Wolsong, Hanbit, Hanul). Wind roses serve as invaluable tool for identifying wind patterns in each region and visualizing wind directions. This can be utilized to predict the dispersion pathway and extent range of radioactive materials carried by the wind. This program will take on the role of establishing appropriate evacuation routes or shelter locations for residents when reliable wind data is not immediately available during an NPP accident. The wind data used in the study was collected from a meteorological tower located at the NPP site, and measurements were taken at 1-hour intervals for each operation over a period of ten years. The collected data underwent preprocessing, followed by the development of Python code to render the wind rose diagrams in an interpretable format. The future direction of this study will be focused on enhancing this program by integrating geographical mapping capabilities. With these advancements, it will become feasible to superimpose shelter positions on a map in accordance with prevailing wind directions. These improvements will contribute to the development of additional protective measures for residents and the proposal of alternative shelter options in response to potential radioactive material releases.

Radiation workers, especially those dealing with Uranium isotopes, can potentially intake Uranium -containing materials through their respiratory and digestive systems. According to the “Regulations on the Measurement and Calculation of Internal Exposure” from Nuclear Safety and Security Commission (NSSC), those who intend to work in or enter the nuclear facilities with a risk of exceeding 2 mSv exposure per year should be examined the internal exposure. However, when it comes to in-vitro bioassay, Uranium intake through drinking water can affect the quantitative analysis. The International Commission on Radiological Protection (ICRP) reported in ICRP Publication 23 (Report on the Task Group on Reference Man) that the reference man excretes Uranium in the urine (0.05-0.5 μg/day) and feces (1.4-1.8 μg/day). Korea Atomic Energy Research Institute (KAERI) set the 90.5 ng/day as the 238U background of workers handing Uranium based on the daily Uranium intake of Koreans. In this research, we examined the possible effects of Uranium in drinking water on internal exposure by analyzing the concentration of Uranium in bottled waters from various water sources sold in the domestic market and a water from the water purifier. The 238U concentration results of analyzing 11 bottled waters and 1 purified water, were ranged from 0 to 10.2 μg/L. All the results were satisfied the standard of 30 μg/L according to “Regulations for Drinking Water Quality Standards and Inspection” enacted by the Ministry of Environment. However, various concentrations were shown depending on the water sources. Assuming that these concentrations of water are consumed by drinking 1 L per day, the internal dose assessment result is 0 to 0.94 mSv. On the other hand, if it is assumed to be inhaled, it can be an overestimated because the dose coefficient of inhalation, Type M is higher than that of ingestion, f1=0.02 which are the values recommended by ICRP Publication 78 (Individual Monitoring for Internal Exposure of Workers) when the Uranium compound is unspecified. In case of two workers at KAERI, the daily excretion of urine was 151 and 120 ng/day respectively in the first quarter monitoring. However after changing the kind of drinking water in the second quarter monitoring, it dropped to 17.4 and 15.4 ng/day respectively. Through this study, it is confirmed that the Uranium background in urine can be analyzed differently depending on the kind of drinking water consumed by each worker. Depending on the Uranium concentration of drinking water, the internal exposure dose assessment can be overestimated or underestimated. Therefore, the Uranium concentration and intake amount according to the kind of drinking water should be considered for in-vitro bioassays of Uranium handlers. Furthermore, if necessary, the Uranium isotope ratio analysis in urine and the handling information should be comprehensively considered. In addition, in order to exclude the effect of intake through the digestive system, replacing the kind of drinking water can be considered. The additional analysis such as in-vivo bioassay and 24 hours urine analysis rather than spot samples can be also recommended.

KEPCO KPS is the contractor for the full system decontamination (FSD) of Kori Unit 1 and under preparation such as modification, lay out for equipment installation, setting up tie-in/out point for chemical injection and way to pressurize the system, of its successful performance. In this research, KPS introduced how KPS has designed and prepared for the FSD project and how will the chemical decontamination process be implemented. As described in the previous research, chemical decontamination process is planned to be conducted for three cycles and each cycle is consisted of oxidation, reduction, decomposition, and purification. Oxidation and reduction process were conducted at 90°C. Chemical decomposition and purification process were conducted at 40°C due to the damage of IX by the heat. If the decontamination result does not meet the target DF and the dose rate, additional cycle can be conducted. Expected volume of process water for FSD is 200 m3. Three systems have been designated as decontamination targets: reactor coolant system (RCS), residual heat removal system (RHRS), chemical volume control system (CVCS). For the steady flow rate, existed plant equipment such as reactor coolant pump (RCP) will be operated and modifications on some components will be conducted. Due to the limited space for installation, decontamination equipment and other resources are distributed to three different places. KPS designed the layout of equipment installed inside the containment vessel. The layout contains the information of shielding for highly radiated equipment such as IX and filter skid.

The domestic Pressurized Heavy Water Reactor (PWHR) nuclear power plant, Wolsong Unit 1, was permanently shut down on December 24, 2019. However, research on decommissioning has mainly focused on Pressurized Water Reactors (PWRs), with a notable absence of both domestic and international experience in the decommissioning of PHWRs. If proper business management such as radiation safety and waste is not performed, it can lead to increased business risks and costs in decommissioning. Therefore, the assessment of waste volume and cost, which provide fundamental data for the nuclear decommissioning process, is a crucial technical requirement before initiating the actual decommissioning of Wolsong Unit 1. Decommissioning radiation-contaminated structures and facilities presents significant challenges due to high radiation levels, making it difficult for workers to access these areas. Therefore, technology development should precede decommissioning process assessments and safety evaluations, facilitating the derivation of optimal decommissioning procedures and ensuring worker safety while enhancing the efficiency of decommissioning operations. In this study, we have developed a program to estimate decommissioning waste amounts for PHWRs, building upon prior research on PWR decommissioning projects while accounting for the specific design characteristics of PHWRs. To evaluate the amount of radioactive waste generated during decommissioning, we considered the characteristics of radioactive waste, disposal methods, packaging container specifications, and the criteria for the transfer of radioactive waste to disposal operators. Based on the derived algorithm, we conducted a detailed design and implemented the program. The proposed program is based on 3D modeling of the decommissioning components and the calculation of the Work Difficulty Factor (WDF), which is used to determine the time weighting factors for each task. Program users can select the cutting and packaging conditions for decommissioning components, estimate waste amount based on the chosen decommissioning method, and calculate costs using time weighting factors. It can be applied not only to PHWRs, but also to PWRs and non-nuclear fields, providing a flexible tool for optimizing decommissioning process.

Chelate resin is a resin that has an exchange group which can form chelates with various metal ions. It shows higher selectivity for metal ions than ion exchange resin and can selectively remove characteristic metal ions. In an aqueous solution containing metal ions, chelate resin can adsorb specific metal ions, and the separated chelate resin can desorb the adsorbed metal ions by changing temperature or pH, so chelate resin has the advantage of being reusable. Chelate resin has been used industrially as an adsorbent to adsorb and separate heavy metal ions in wastewater, and is also used for the purpose of recovering precious or rare metals contained in industrial wastewater or industrial waste. Against this background, there is a need to develop chelate resins with higher adsorption capacity. Acrylic fiber is defined as a man-made fiber made from a linear synthetic polymer with fiberforming ability consisting of more than 85% acrylonitrile. It is a man-made fiber that is often used as a substitute for wool because it has good thermal insulation properties like wool and is warm and soft to the touch. It is a fiber rich in cyano groups due to its high content of acrylonitrile, and has the advantage of being able to be used as a variety of functional fibers through modification of cyano groups. In this study, the amination reaction of acrylic fiber was performed using diethylenetriamine, and the adsorption characteristics for metal ions were evaluated according to the reaction conversion rate. In order to improve the amination efficiency, 400 kGy was irradiated using a 2.5 MeV electron beam accelerator, and through this, the crosslinking rate of acrylic fiber was able to be improved up to 80%. Water and ethanol were used as cosolvents for the amination reaction in a ratio of 60/40 vol/vol, respectively, and a reaction yield of 178% was obtained after 120 minutes of reaction. Using the chelate resin prepared in this way, the adsorption performance for metal ions was evaluated through Atomic Absorption Spectrometry analysis.

Radioactive contamination distribution in nuclear facilities is typically measured and analyzed using radiation sensors. Since generally used detection sensors have relatively high efficiency, it is difficult to apply them to a high radiation field. Therefore, shielding/collimators and small size detectors are typically used. Nevertheless, problems of pulse accumulation and dead time still remain. This can cause measurement errors and distort the energy spectrum. In this study, this problem was confirmed through experiments, and signal pile-up and dead time correction studies were performed. A detection system combining a GAGG sensor and SiPM with a size of 10 mm × 10 mm × 10 mm was used, and GAGG radiation characteristics were evaluated for each radiation dose (0.001~57 mSv/h). As a result, efficiency increased as the dose increased, but the energy spectrum tended to shift to the left. At a radiation dose intensity of 400 Ci (14.8 TBq), a collimator was additionally installed, but efficiency decreased and the spectrum was distorted. It was analyzed that signal loss occurred when more than 1 million particles were incident on the detector. In this high-radioactivity area, quantitative analysis is likely to be difficult due to spectral distortion, and this needs to be supplemented through a correction algorithm. In recent research cases, the development of correction algorithms using MCNP and AI is being actively carried out around the world, and more than 98% of the signals have been corrected and the spectrum has been restored. Nevertheless, the artificial intelligence (AI) results were based on only 2-3 overlapping pulse data and did not consider the effect of noise, so they did not solve realistic problems. Additional research is needed. In the future, we plan to conduct signal correction research using ≈10×10 mm small size detectors (GAGG, CZT etc.). Also, the performance evaluation of the measurement/analysis system is intended to be performed in an environment similar to the high radiation field of an actual nuclear facility.

Activated carbon (AC) is used for filtering organic and radioactive particles, in liquid and ventilation systems, respectively. Spent ACs (SACs) are stored till decaying to clearance level before disposal, but some SACs are found to contain C-14, a radioactive isotopes 5,730 years halflife, at a concentration greater than clearance level concentration, 1 Bq/g. However, without waste acceptance criteria (WAC) regarding SACs, SACs are not delivered for disposal at current situation. Therefore, this paper aims to perform a preliminary disposal safety examination to provide fundamental data to establish WAC regarding SACs SACs are inorganic ash composed mostly of carbon (~88%) with few other elements (S, H, O, etc.). Some of these SACs produced from NPPs are found to contain C-14 at concentration up to very-low level waste (VLLW) criteria, and few up to low-level waste (LLW) criteria. As SACs are in form of bead or pellets, dispersion may become a concern, thus requiring conditioning to be indispersible, and considering VLL soils can be disposed by packaging into soft-bags, VLL SACs can also be disposed in the same way, provided SACs are dried to meet free water requirement. But, further analysis is required to evaluate radioactive inventory before disposal. Disposability of SACs is examined based on domestic WAC’s requirement on physical and chemical characteristics. Firstly, particulate regulation would be satisfied, as commonly used ACs in filters are in size greater than 0.3 mm, which is greater than regulated particle size of 0.2 mm and below. Secondly, chelating content regulation would be satisfied, as SACs do not contain chelating chemicals. Also, cellulose, which is known to produce chelating agent (ISA), would be degraded and removed as ACs are produced by pyrolysis at 1,000°C, while thermal degradation of cellulose occurs around 350~600°C. Thirdly, ignitability regulation would be satisfied because as per 40 CFR 261.21, ignitable material is defined with ignition point below 60°C, but SACs has ignition point above 350°C. Lastly, gas generation regulation would be satisfied, as SACs being inorganic, they would be targeted for biological degradation, which is one of the main mechanism of gas generation. Therefore, SACs would be suitable to be disposed at domestic repositories, provided they are securely packaged. Further analysis would be required before disposal to determine detailed radioactive inventories and chemical contents, which also would be used to produce fundamental data to establish WAC.

The development of existing radioactive waste (RI waste) management technologies has been limited to processing techniques for volume reduction. However, this approach has limitations as it does not address issues that compromise the safety of RI waste management, such as the leakage of radioactive liquid, radiation exposure, fire hazards, and off-gas generation. RI waste comes in various forms of radioactive contamination levels, and the sources of waste generation are not fixed, making it challenging to apply conventional decommissioning and disposal techniques from nuclear power plants. This necessitates the development of new disposal facilities suitable for domestic use. Various methods have been considered for the solidification of RI waste, including cement solidification, paraffin solidification, and polymer solidification. Among these, the polymer solidification method is currently regarded as the most suitable material for RI waste immobilization, aiming to overcome the limitations of cement and paraffin solidification methods. Therefore, in this study, a conceptual design for a solidification system using polymer solidification was developed. Taking into account industrial applicability and process costs, a solidification system using epoxy resin was designed. The developed solidification system consists of a pre-treatment system (fine crush), solidification system, cladding system, and packing system. Each process is automated to enhance safety by minimizing user exposure to radioactive waste. The cladding system was designed to minimize defects in the solidified material. Based on the proposed conceptual design in this paper, we plan to proceed with the specific design phase and manufacture performance testing equipment based on the basic design.

Domestic waste acceptance criteria (WAC) require flowable or homogeneous wastes, such as spent resin, concentrated waste, and sludge, etc., to be solidified regardless of radiation level, to provide structural integrity to prevent collapse of repository, and prevent leaching. Therefore, verylow level (VLL) spent resin (SR) would also require to be solidified. However, such disposal would be too conservative, considering IAEA standards do not require robust containment and shielding of VLL wastes. To prevent unnecessary cost and exposure to workers, current WAC advisable to be amended, thus this paper aims to provide modified regulation based on reviewed engineering background of solidification requirement. According to NRC report, SR is classified as wet-solid waste, which is defined as a solid waste produced from liquid system, thus containing free-liquid within the waste. NRC requires liquid wastes to be solidified regardless of radiation level to prevent free liquid from being disposed, which could cause rapid release of radionuclides. Furthermore, considering class A waste does not require structural integrity, unlike class B and C wastes, dewatering would be an enough measure for solidification. This is supported by the cases of Palo Verde and Diablo Canyon nuclear power plants, whose wet-solid wastes, such as concentrated wastes and sludge, are disposed by packaging into steel boxes after dewatering or incineration. Therefore, dewatering VLL spent resin and packaging them into structural secure packaging could satisfy solidification goal. Another goal of solidification is to provide structural support, which was considered to prevent collapse of soil covers in landfills or trenches. However, providing structural support via solidification agent (ex. Cement) would be unnecessary in domestic 2nd phase repository. As the domestic 2nd phase repository is cementitious structure, which is backfilled with cement upon closure, the repository itself already has enough structural integrity to prevent collapse. Goldsim simulation was run to evaluate radiation impact by VLL SR, with and without solidification, by modelling solidified wastes with simple leaching, and unsolidified wastes with instant release. Both simulations showed negligible impact on radiation exposure, meaning that solidifying VLL SR to delay leaching would be irrational. Therefore, dewatering VLL SR and packaging it into a secure drum (ex. Steel drum) could achieve solidification goals described in NRC reports and provide enough safety to be disposed into domestic repositories. In future, the studied backgrounds in this paper should be considered to modify current WAC to achieve efficient waste management.

EU taxonomy requires to solve problems for safe management of radioactive waste and disposal of spent fuel, which is a precondition for growing demand for nuclear power plant. Currently, Korea manages about 18,000 tons of high-level radioactive waste at temporary storage facilities in nuclear power plant sites, but such temporary storage facilities are expected to become saturated sequentially from 2031. Therefore, it is necessary to secure a permanent disposal facility to safely treat high-level radioactive waste. In accordance with the second basic plan for high-level radioactive waste management in 2021, it is necessary to establish requirements for regulatory compliance for the site selection and site acquisition, investigation and evaluation, and construction for the establishment of a deep geological disposal facility. In this study, we analyzed the regulatory policies and cases of leading foreign countries related to deep geological disposal facilities for high-level radioactive waste disposal waste such as IAEA, USA, Sweden, and Finland using data analysis methodology. To analyze a large amount of textbased document data, text mining is applied as a major technology and a verification standard that secures validity and safety based on the regulatory laws described so far is developed to establish a regulatory base suitable for domestic deep geological disposal status. Based on the collected data, preprocessing and analysis with Python were performed. Keywords and their frequency were extracted from the data through keyword analysis. Through the measured frequency values, the contents of the objects and elements to be regulated in the statutory items were grasped. And through the frequency values of words co-occurring among different sections through the analysis of related words, the association was obtained, and the overall interpretation of the data was performed. The results of analyzing regulations of major foreign countries using text mining are visualized in charts and graphs. Word cloud can intuitively grasp the contents by extracting the main keywords of the contents of the regulations. Through the network connection graph, the relationship between related words can be visually structured to interpret data and identify the causal relationship between words. Based on the result data, it is possible to compare and analyze the factors to be supplemented by analyzing domestic nuclear safety case and regulations.

It has been known that as oxide layer (ZrO2) forms on the nuclear fuel cladding during irradiation in nuclear power plants, the corrosion kinetics are influenced by various parameters such as chemical environments. One of those environments, crud deposition driven by coolant chemistry has an adverse effect on the formation of oxide (ZrO2) and leads to increase thickness of the layer. In this study, crud formation was performed through loop experiment equipment on the surface of intentionally-made oxide layer (ZrO2) on cladding tubes and then the composition and characteristics of cruds were examined for the investigation of nuclear power plant environment. As a result, various cruds in composition and microstructure were formed depending on the exquisite methods and conditions such as metal ion concentration.

In nuclear facilities, a graded approach is applied to achieve safety effectively and efficiently. It means that the structures, systems, and components (SSCs) that are important to safety should be assured to be high quality. Accordingly, SSCs that consist of nuclear facilities should be classified with respect to their safety importance as several classes, so that the requirements of quality assurance relevant to the designing, manufacturing, testing, maintenance, etc. can be applied. Guidance for the safety classification of SSCs consisting of nuclear power plants and radioactive waste management facilities was developed by U.S.NRC and IAEA. Especially, in guidance for nuclear power plants, safety significance can be evaluated as following details. The single SSC that mitigates or/and prevents the radiological consequence or hazard was assumed to be failure or malfunction as the initiating event/accident occurred and the following radiological consequence was evaluated. Considering both the consequence and frequency of the occurrence of the initiating event/accident, the safety significance of each SSC can be evaluated. Based on the evaluated safety significance, a safety class can be assigned. The guidance for the safety classification of the spent nuclear fuel dry storage systems (DSS) was also developed in the United States (NUREG/CR-6407) and the U.S.NRC acknowledges the application of it to the safety classification of DSS in the United States. Also, worldwide including the KOREA, that guidance has been applied to several DSSs. However, the guidance does not include the methodology for classifying the safety or the evaluated safety significance of each SSC, and the classification criteria are not based on quantitative safety significance but are expressed somewhat qualitatively. Vendors of DSS may have difficulties to apply this guidance appropriately due to the different design characteristics of DSSs. Therefore, the purpose of this study is to evaluate the safety significance of representative SSCs in DSS. A framework was established to evaluate the safety significance of SSCs performing safety functions related to radiation shielding and confinement of radioactive materials. Furthermore, the framework was applied to the test case.

The Nuclear Export and Import Control System (NEPS) is currently in operation for nuclear export and import control. To ensure consistent and efficient control, various computational systems are either already in place or being developed. With numerous scattered systems, it becomes crucial to integrate the databases from each to maximize their utility. In order to effectively utilize these scattered computer systems, it is necessary to integrate the databases of each system and develop an associated search system that can be used for integrated databases, so we investigated and analyzed the AI language model that can be applied to the associated search system. Language Models (LM) are primarily divided into two categories: understanding and generative. Understanding Language Models aim to precisely comprehend and analyze the provided text’s meaning. They consider the text’s bidirectional context to understand its deeper implications and are used in tasks such as text classification, sentiment analysis, question answering, and named entity recognition. In contrast, Generative Language Models focus on generating new text based on the given context. They produce new textual content continuously and are beneficial for text generation, machine translation, sentence completion, and storytelling. Given that the primary purpose of our associated search system is to comprehend user sentences or queries accurately, understanding language models are deemed more suitable. Among the understanding language models, we examined BERT and its derivatives, RoBERTa and DeBERTa. BERT (Bidirectional Encoder Representations from Transformers) uses a Bidirectional Transformer Encoder to understand the sentence context and engages in pre-training by predicting ‘MASKED’ segments. RoBERTa (A Robustly Optimized BERT Pre-training Approach) enhances BERT by optimizing its training methods and data processing. Although its core architecture is similar to BERT, it incorporates improvements such as eliminating the NSP (Next Sentence Prediction) task, introducing dynamic masking techniques, and refining training data volume, methodologies, and hyperparameters. DeBERTa (Decoding-enhanced BERT with disentangled attention) introduces a disentangled attention mechanism to the BERT architecture, calculating the relative importance score between word pairs to distribute attention more effectively and improve performance. In analyzing the three models, RoBERTa and DeBERTa demonstrated superior performance compared to BERT. However, considering factors like the acquisition and processing of training data, training time, and associated costs, these superior models may require additional efforts and resources. It’s therefore crucial to select a language model by evaluating the economic implications, objectives, training strategies, performance-assessing datasets, and hardware environments. Additionally, it was noted that by fine-tuning with methods from RoBERTa or DeBERTa based on pre-trained BERT models, the training speed could be significantly improved.

The National R&D Innovation Act emphasizes the improvement of the quality of R&D activities. The research institute is making efforts to improve the quality of research and effectively manage research implementation. KINAC has conducted various R&D projects regarding nuclear nonproliferation and nuclear security, and their scope and scale have been gradually more widened and increased. It consequently becomes important how to successfully manage research projects and ensure their qualification with the growth and complexity of research in KINAC. Unfortunately, no attempt was made to introduce and apply project management methodologies. Therefore, the objective of this study is to introduce project management standards and guidelines as an initial step towards improving the overall research quality of the institute. Project management is the well-organized application of knowledge and techniques to efficiently and effectively initiate, plan, control, and close projects, in order to achieve specific goals and meet success criteria. There are some guidelines regarding project management, including PMBOK (the Project Management Body of Knowledge), PRINCE2 (Projects in Controlled Environments), ISO 21500 (Guidance on Project Management), and PMP (Project Management Professionals), etc. They are international standards that consist of processes, guidelines, and best practices for project management. They provide structured processes and approaches to plan, execute, monitor, control, and complete projects. By reviewing the guidelines, the commonly important factors, including schedule, cost, quality, resources, communication, and risk management were introduced to apply to KINAC R&D project implementation. In addition to the management standards, systematic efforts are also continued to enhance the R&D qualities of the institute. These efforts include the implementation of a quality management system (ISO 9001:2015), development of an integrated research achievements management system, regulation development, and distribution of guidebooks for project managers and researchers. These efforts have been evaluated as improving the quality of the research.

ISO 9001:2005 is the international standard for implementing a Quality Management System (QMS), which provides a framework and principles for managing an organization’s quality management. The aim is to ensure that the organization continuously provides products and services that satisfy regulatory requirements. The “process approach” in ISO 9001 is defined as a systematic method of achieving organizational goals by comprehending and managing the interconnected processes as a cohesive system. Recently, KINAC has decided to develop standard processes in the field of R&D and performance management based on the framework of the ISO 9001:2015 quality management system. The objective of this study is to establish standardized processes for conducting research and development, as well as managing the outputs and performance of R&D activities. It involves identifying, designing, implementing, monitoring, and continually improving processes to ensure consistency, efficiency, and effective management of KINAC R&D and its achievements. Firstly, R&D and the research performance management process were defined, and the processes were categorized by function according to the requirements of ISO 9001:2015. Second, the ISO 9001 requirements were compared to the institute’s existing regulations and documents in order to identify any additional processes and procedures needed to meet the quality management requirements. Finally, the lists of quality documentation were determined for the institute’s QMS. As a result, a total of 30 QMS documents were listed, including 1 manual, 12 quality processes and procedures, and 17 quality instructions. The documents can be categorized into four process groups: the management and planning process group, the R&D and achievements management process group, the analysis and improvement process group, and the support process group. All input and output information of each process are connected and interrelated. The implementation of quality management standards and procedures for R&D in KINAC could lead to improved research practices, more reliable data collection and analysis, and increased efficiency in conducting R&D activities. For further study, it is planned to create detailed, high-quality documents that adhere to standard requirements and guidelines.

The Republic of Korea (ROK), as a member state of the IAEA, is operating the State’s System of Accounting for and Control (SSAC) and conducting independent national inspections. Furthermore, an evaluation methodology for the material unaccounted for (MUF) is being developed in ROK to enhance capabilities of national inspection. Generally, physical and chemical changes of nuclear material are unavoidable due to the operating system and structure of facilities, an accumulation of material unaccounted for (MUF) has been issued. IAEA developed statistical MUF evaluation method that can be applied to all facilities around the world and it mainly focuses on the diversion detection of nuclear materials in facilities. However, in terms of the national safeguard inspection, an evaluation of accountancy in facilities is additionally needed. Therefore, in this research, a new approach to MUF evaluation is suggested, based on the Guide to the Expression of Uncertainty in Measurement (GUM) that an evaluation of measurement uncertainty factors is straightforward. A hypothetical list of inventory items (LII) which has 6,118 items at the beginning and end of the material balance period, along with 360 inflow and outflow nuclear material items at a virtual fuel fabrication plant was employed for both the conventional IAEA MUF evaluation method and the proposed GUM-based method. To calculate the measurement uncertainty, it was assumed that an electronic balance, gravimetry, and a thermal ionization mass spectrometer were used for a measurement of the mass, concentration, and enrichment of 235U, respectively. Additionally, it was considered that independent and correlated uncertainty factors were defined as random factors and systematic factors for the ease of uncertainty propagation by the GUM. The total MUF uncertainties of IAEA (σMUF) and GUM (uMUF) method were 37.951 and 36.692 kg, respectively, under the aforementioned assumptions. The difference is low, it was demonstrated that the GUM method is applicable to the MUF evaluation. The IAEA method demonstrated its applicability to all nuclear facilities, but its calculated errors exhibited low traceability due to its simplification. In contrast, the calculated uncertainty based on the GUM method exhibited high reliability and traceability, as it allows for individual management of measurement uncertainty based on the facility’s accounting information. Consequently, the application of the GUM approach could offer more benefits than the conventional IAEA method in cases of national safeguard inspections where factor analysis is required for MUF assessment.

Spodoptera litura is an omnivorous pest that damages more than 120 types of crops. It is widely distributed throughout the tropics, subtropics, and temperate zones. It is a major pest that causes damage by explosively increasing in density in a short period of time due to its excellent mobility and reproductive ability. S. litura was studied at five constant temperatures (15.0, 20.0, 25.0, 30.0, and 35.0℃), 65±5% relative humidity (RH), and 14L:10D photoperiods. The development period for eggs and larvae tended to become shorter as temperature increased. The growth period of egg, larva, and pupa by 25℃ temperature was 4.0, 32.3, and 12.6 days, respectively.

마늘은 백합과 알리움속에 속하는 영양체작물로 누대재배시 수확량이 감소하는 특성을 보인다. 마늘의 화경 끝에는 주아가 생성되는데, 주아를 파종하여 2년이상 누대재배시 수확량이 약 15% 이상 증가하기 때문에 주아재 배를 통한 종구 갱신이 필요하다. 마늘 주아를 수확 후 저장시 파총채벌레, 파좀나방 등이 발생하며, 저장 중 주아 는 매년 평균 20% 이상의 부패한다. 주아재배를 원활히 하기 위해 주아 저장 시 해충의 방제를 통해 부패율을 줄이는 것이 중요하다. 본 연구에서 주아에서의 해충 발생밀도 및 방제방법에 대한 연구를 진행하였다. 연구결과 주아에서 파총채벌레는 7월 3주차, 파좀나방은 7월 2주차에 각각 발생최성기를 보였으며, 약제 침지시 방제가가 90% 이상 관찰되었다. 약제 침지된 주아 파종시 약해는 발생하지 않아, 주아 해충 방제시 침지법을 활용하여 해충의 밀도를 조절 할 수 있을 것으로 판단된다.

마늘은 백합과 알리움속에 속하는 작물로 약용, 양념채소로 많이 소비되어져 왔다. 마늘은 6월에 수확하여 1개월 간의 건조과정을 거치게 되는데 보통 장마기와 겹치게 되어 연평균 부패율은 5%이상으로 높다. 마늘의 저장중 병해충은 뿌리응애, 마늘혹응애, 마른썩음병 등으로 섭식에 따른 부패를 유발하며, 건조시기를 단축하고 효율적인 건조를 통하여 감모율을 줄이는 것이 중요하다. 마늘의 병해충 피해를 줄이기 위해 개발한 열풍 흡기식 건조장치를 포함하여 관행, 열풍 건조기, 흡기식 건조 등 4가지 방법으로 마늘을 건조하였으며, 방법별 건조소요 일수, 부패율 등을 조사하였다. 연구결과 열풍-흡기식 건조장치의 순환 공기의 온습도는 외기대비 7.8℃ 높았고, 28.6% 낮았으며, 건조 소요일수는 관행건조 대비 31% 수준으로 우수하였다. 또한 병해충 피해 양상은 관행 건조 대비 4.4%p낮아 건조 기간 및 정상품율이 관행 대비 우수하여, 관행 건조를 대체 가능할 것으로 생각된다.