In this study, we examined the antagonistic effects of sprout-borne lactic acid bacteria (LAB) on Salmonella enterica serovar Enteritidis. This antagonism is promoted as a means of controlling contamination during sprout production and provides additional LAB for consumers. We isolated a total of 24 LAB isolates in nine species and five genera from seven popular vegetable sprouts: alfalfa (Medicago sativa), clover (Trifolium pratense), broccoli (Brassica oleracea ssp. italica), vitamin (B. rapa ssp. narinosa), red radish (Raphanus sativus), red kohlrabi (B. oleracea var. gongylodes), and Kimchi cabbage (B. campestris var. pekinensis). Based on 16S rRNA gene sequences, the LAB species were identified as Enterococcus casseliflavus, E. faecium, E. gallinarum, E. mundtii, Lactococcus taiwanensis, Leuconostoc mesenteroides, Pediococcus pentosaceus, and Weissella cibaria, and W. confusa. A total of 16 LAB isolates in seven species including E. faecium, E. gallinarum, E. mundtii, L. taiwanensis, L. mesenteroides, P. pentosaceus, and W. cibaria showed antagonistic activity toward S. enterica. The growth inhibition of sprout LAB on S. enterica was confirmed by co-culture. Unexpectedly, sprout LAB failed to suppress the growth of S. enterica in alfalfa sprouts, whereas all LAB strains stimulate S. enterica growth even if it is not significant in some strains. The findings of this study indicate that S. enterica-antagonistic LAB are detrimental to food hygiene and will contribute to further LAB research and improved vegetable sprout production.

When the parent radionuclide decays, the progeny radionuclide is produced. Accordingly, the dose contribution of the progeny radionuclide should be considered when assessing dose. For this purpose, European Commission (EC) and International Atomic Energy Agency (IAEA) provide weighting factors for dose coefficient. However, these weighting factors have a limitation that does not reflect the latest nuclide data. Therefore, in this study, we analyzed the EC and IAEA methodology for derivation of weighting factor and used the latest nuclide data from ICRP 107 to derive weighting factors for dose coefficient. Weighting factor calculation is carried out through 1) selection of nuclide, 2) setting of evaluation period, and 3) derivation based on ICRP 107 radionuclide data. Firstly, in order to derive the weighting factor, we need to select the radionuclides whose dose contribution should be considered. If the half-life of progeny radionuclides sufficiently short compared to the parent radionuclide to achieve radioactive equilibrium, or if the dose coefficient is greater of similar to that of the parent radionuclide and cannot be ignored, the dose contribution of the progeny radionuclide should be considered. In order not to underestimate the dose contribution of progeny radionuclides, the weighting factors for the progeny nuclides are taken as the maximum activity ratio that the respective progeny radionuclides will reach during a time span of 100 years. Finally, the weighting factor can be derived by considering the radioactivity ratio and branch fraction. In order to calculate the weighting factor, decay data such as the half-life of the radionuclide, decay chain, and branch fraction are required. In this study, radionuclide data from ICRP 107 was used. As a result of the evaluation, for most radionuclides, the weighting factors were derived similarly to the existing EC and IAEA weighting factors. However, for some nuclides, the weighting factors were significantly different from EC and IAEA. This is judged to be a difference in the half-life and branch fraction of the radionuclide. For example, in the case of 95Zr, the weighting factor for 95mNb showed a 35.8% difference between this study and previous study. For ICRP 38, when 95Zr decays, the branch fraction for 95mNb is 6.98×10-3. In contrast, for ICRP 107, the branch fraction is 1.08×10-2, a difference of 54.7%. Therefore, the weighting factor for the dose coefficient based on ICRP 107 data may differ from existing studies depending on the half-life and decay information of the nuclide. This suggests the need for a weighting factor based on the latest nuclide data. The results of this study can be used as a basis for the consideration of dose contributions for progeny radionuclides in various dose assessments.

The demand for transportation is increasing due to the continuous generation of radioactive wastes. Especially, considering the geographical characteristics of Korea and the location characteristics of nuclear facilities, the demand for maritime transportation is expected to increase. If a sinking accident happens during maritime transportation, radioactive materials can be released into the ocean from radioactive waste transportation containers. Radioactive materials can spread through the ocean currents and have radiological effects on humans. The effect on humans is proportional to the concentration of radioactive materials in the ocean compartment. In order to calculate the concentration of radioactive materials that constantly flow along the ocean current, it is necessary to divide the wide ocean into appropriate compartments and express the transfer processes of radioactive materials between the compartments. Accordingly, this study analyzed various ocean transfer evaluation methodologies of overseas maritime transportation risk codes. MARINRAD, POSEIDON, and LAMER codes were selected to analyze the maritime transfer evaluation methodology. MARINRAD divided the ocean into two types of compartments that water and sediment compartments. And it was assumed that radionuclides are transfered from water to water or from water to sediment. Advection, diffusion, and sedimentation were established as transfer process for radionuclides between compartments. MARINRAD use transfer parameters to evaluate transer processes by advection, diffusion, and sedimentation. Transfer parameters were affected by flow rate, sedimentation rate, sediment porosity, and etc. POSEIDON also divided the ocean into two types that water and sediment compartment, each compartments was detaily divided into three vertical sub-compartment. Advection, diffusion, resuspension, sedimentation, and bioturbation were established as transport processes for radionuclides between compartments. POSEIDON also used transfer parameters for evaluating advection, diffusion, resuspension, sedimentation, and bioturbation. Transfer parameters were affected by suspended sediment rates, sedimentation rates, vertical diffusion coefficients, bioturbation factors, porosity, and etc. LAMER only considered the water compartment. It divided the water compartment into vertical detailed compartments. Diffusion, advection and sedimentation were established as the nuclide transfer processes between the compartments. To evaluated the transfer processes of nuclides for diffusion and advection, LAMER calculated the probability with generating random position vectors for radionuclides’ locations rather than deterministic methods such as MARINRAD’s transfer parameters or POSEIDON’s transfer rates to evaluate transfer processes. The results of this study can be used as a basis for developing radioactive materials’ ocean transfer evaluation model.

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.

The operation of nuclear facilities involves the potential for on-site contamination of soil, primarily resulting from pipe leaks and other operational incidents. Globally, decommissioning process for commercial nuclear power plants have revealed huge-amounts of soil waste contaminated with Cs-137, Sr-90, Co-60, and H-3. For example, Connecticut Yankee in the United States produced approximately 52,800 ton of contaminated soil waste, constituting 10% of the total waste generated during its decommissioning. Environmental remediation costs associated with nuclear decommissioning in the US averaged $60 million per unit, representing a significant 10% of the whole decommissioning expenses. Consequently, this study undertook a preliminary investigation to identify important factors for establishing a site remediation strategy based on radionuclide- and site-specific media- characteristics, focusing the efficiency enhancement for the environmental remediation. The factors considered for this investigation were categorized into physical/environmental, socioeconomic, technical, and management aspects. Physical/environmental factors contained the site characteristics, contamination levels, and environmental sensitivity, while socio-economic factors included the social concerns and economic costs. Technical and management factors included subcategories such as technical considerations, policy aspects, and management factors. Especially, technical factors were further subdivided to consider the site reuse potential, secondary waste generation by site remediation, remediation efficiency, and remediation time. Additionally, our study focused the key factors that facilitate the systematic planning for the site remediation, considering the distribution coefficient (Kd) and hydrogeological characteristics associated with each radionuclide in specific site conditions. Therefore, key factors in this study focus the geochemical characteristics of site media including the particle size distribution, chemical composition, organic and inorganic constituents, and soil moisture content. Moreover, the adsorption properties of site media were examined concerning the distribution coefficient (Kd) of radionuclides and their migration characteristics. Furthermore, this study supported the development of a conceptual framework, containing the remediation strategies that incorporate the mobility of radionuclides, according to the site-specific media. This conceptual framework would necessitate the spatial analysis techniques involving the whole contamination surveys and radionuclide mobility modeling data. By integrating these key factors, the study provides the selection and simulation of optimal remediation methods, ultimately offering the estimated amounts of radioactive waste and its disposal costs. Therefore, these key factors offer foundational insights for designing the site remediation strategies according the sitespecific information such as the distribution coefficient (Kd) and hydrogeological characteristics.

Hydrogen isotope separation involves the separation of hydrogen, deuterium, tritium, and their isotopologues. It is an essential technology for removing radioactive tritium contamination and for obtaining valuable hydrogen isotope resources. Among various hydrogen isotope separation technologies, water electrolysis technology exhibits a high separation factor. Consequently, the electrolysis of tritiated water is of paramount importance as a tritium enrichment method for treating tritium-contaminated water and for analyzing tritium in environmental samples. More recently, hydroelectrolysis technology, which utilizes proton exchange membranes (PEM) to reduce water inventory, has gained favor over traditional alkaline hydroelectrolysis. Nevertheless, it is crucial to decrease the hydrogen permeability of the PEM in order to mitigate the explosion risk associated with tritium hydrogen electrolysis devices. Additionally, efforts are needed to enhance the hydrogen isotope selectivity of the PEM and optimize the manufacturing process of the membrane-electrode assembly (MEA), thereby improving both hydrogen isotope separation performance and water electrolysis efficiency. In this presentation, we will delve into two key aspects. Firstly, we’ll explore the reduction of hydrogen permeability and the enhancement of the hydrogen isotope separation factor in PEM through the incorporation of 2D nanomaterial additives. Secondly, we’ll examine the influence of various MEAs preparation methods on electrolysis and isotope separation performances. Lastly, we will discuss the effectiveness of the developed system in separating deuterium and tritium.

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.

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.

In Korea, most temporary storage facilities for spent nuclear fuel are nearing saturation. As an alternative to this, the 2nd basic plan for high-level radioactive waste management specified the operation plan of dry interim storage facility. Meanwhile, the NSSC No. 2021-19 stipulates that it is necessary to evaluate the possibility and potential effect of accident before operating interim storage facility. Therefore, this study analyzed the categories of accident scenarios that may occur in dry storage facility as part of prior research on this. We investigated the case of categorization of dry storage facility accident scenarios of IAEA, NRC, KAREI, and KINS. The IAEA presented accident scenarios that could occur in on-site dry storage facility operated with silo and cask method. NRC has classified accident scenarios in dry storage facility and estimated the probability of accidents for each. KAERI and KINS selected major accident scenarios and analyzed the processes for each, in preparation for the introduction of dry storage facility in Korea in the future. Overall, a total of 10 accident scenarios were considered, and the scenarios considered by each institution were different. Among 10 scenarios, cask drop and aircraft collision were included in the categorization of most institutions. The results of this study can be used as basic data for cataloging accidents subject to safety evaluation when introducing dry interim storage facility in Korea in the future.

Sweet pepper(paprika) belongs to the genus Capsicum, and is one of the most important export product from Korea to Japan and Southeast Asia. So it is important to eradicate plant quarantine pests before export sweet pepper. Aphids, whiteflies and mites are major pests that can damage to sweet peppers. Fumigation is normally used to eradicate pests in plant quarantine, but phytotoxicity may can be appeared that affect the quality of the product. Low-temperature treatment, one of the most popular physical treatment, can reduce crop damage to preserve product quality, but it takes long time to kill pests, which can cause quality degradation. In this study, phytotoxicity of fumigants, phosphine(PH3), ethyl formate(EF) and PH3+EF on sweet peppers was investigated to use as basic data for physicochemical treatment. When treated with more than 35 mg/L of EF, phytotoxicity was occurred, and was not occurred with PH3. When low-temperature of 1.7 degrees treated for 15 days after fumigation, it seems to be no direct damage from low-temperature treatment. But quality of top of sweet pepper was decreased from 7 days after fumigation.

지난 2022년 제주도 애월읍 일대에서 콩 해충으로 알려진 콩은무늬밤나방(Ctenoplusia agnata) 성충이 검거세 미밤나방(Agrotis ipsilon)의 성페로몬 트랩에 대량으로 포획되었다. 검거세미밤나방 트랩은 목적 해충에 대한 포획 효율을 조사하기 위해 세 구성 성분, (Z)-7-dodecenyl acetate, (Z)-9-tetradecenyl acetate를 3:1 비율로 고정하고 (Z)-11-hexadecenyl acetate를 0, 1, 6, 10, 15로 각각 비율을 달리한 미끼를 사용하였다. 각 조성별 콩은무늬밤나방 성충 포획수를 비교한 결과, (Z)-11-hexadecenyl acetate가 첨가되지 않은 트랩에서 주당 평균 약 17.96마리로, 가장 많은 수의 개체가 포획된 것으로 확인되었다. 반면, (Z)-11-hexadecenyl acetate가 가장 많이 함유된 트랩에서 주당 평균 약 2.5마리로 가장 적은 개체가 포획된 것으로 파악되었다. 이에 (Z)-11-hexadecenyl acetate의 비율이 증가할 수록, 포획되는 콩은무늬밤나방의 개체 수가 감소되는 것을 확인할 수 있었다. 검거세미밤나방 미끼의 주성분인 Z)-7-dodecenyl acetate는 기존의 콩은무늬밤나방 유인 성분 중 하나이기도 하여 해당 성분의 구성비가 유인에 영향을 미쳤다는 것을 예측할 수 있다. 추후에 해당 트랩들과 시판 중인 콩은무늬밤나방 성페로몬 트랩을 설치하 여 포획 양상을 비교할 필요성이 요구된다.

Bombyx mandarina (Lepidoptera: Bombycidae), the presumed ancestor of B. mori, has long been a subject of study to illustrate the geographic relationships in connection with origin of B. mori. We report 97 mitochondrial genome (mitogenome) sequences of B. mandarina collected from Korea and Japan. Phylogenetic and population genetic analyses showed that all individuals of B. mandarina collected in Korean localities formed a strong group together with all individuals originated from northern China (mainly north of the Qinling-Huaihe line) and some of southern China. This group was placed as the sister group to B. mori strians suggesting that this group had been served as an immediate progenitor for B. mori.

해충에 이용되는 화학적 기피제는 생태계를 파괴할 수 있으며 내성을 가진 생물체로의 진화를 촉진한다. 같은 종의 생물끼리의 의사소통 수단인 페로몬을 이용하면 다른 종에게 영향을 미치지 않으면서 특정 곤충에 특이적 으로 작용하는 방충제를 제작할 수 있을 것이라 생각된다. 본 연구는 초파리(Drosophila)의 페로몬 2종류를 추출 하여 초파리의 기피도 및 유인도와 번식률을 확인하고자 한다. ℃, 광주기 12h/12h의 동일한 조건에서 사육하 며 10마리당 헥세인 10를 사용하여 암컷의 표피에서 CHC 페로몬과 수컷의 페로몬샘에서 cVA 페로몬을 추출 한다. 연령, 성별, 교배 여부에 따라 관찰통에 각각의 페로몬을 처리하여 지정구간에 분포하는 초파리의 수를 계수하여 기피도 및 유인도를 확인한다. 관병에 암수 1쌍을 투입하고 하루에 1번 선정한 페로몬을 투여하며 산란 수을 측정한다. 이 연구를 통해 CHC가 수컷 초파리에 대한 기피 효과가 있음을 확인하였으며 추출되는 수컷의 연령이 높을수록 cVA에 의한 번식률 감소가 크게 나타났다. 본 연구를 통해 페로몬을 통한 초파리의 방제 가능성 을 확인하였으므로 다른 곤충의 방제에도 적용할 수 있을 거라 기대한다. 페로몬은 생물 농축과 같은 환경적 영향이 없으며 소량으로 유의미한 결과를 도출했다는 점에서 의의가 있으며 상용화를 통해 해충에 의해 피해를 해결할 수 있을 것이라 기대한다.

This study was aimed to isolate bacterial inoculants producing chitinase and evaluate their application effects on corn silage. Four corn silages were collected from four beef cattle farms to serve as the sources of bacterial inoculants. All isolates were tested against Fusarium graminearum head blight fungus MHGNU F132 to confirm their antifungal effects. The enzyme activities (carboxylesterase and chitinase) were also measured to isolate the bacterial inoculant. Based on the activities of anti-head blight fungus, carboxylesterase, and chitinase, L. buchneri L11-1 and L. paracasei L9-3 were subjected to silage production. Corn forage (cv. Gwangpyeongok) was ensiled into a 10 L mini silo (5 kg) in quadruplication for 90 days. A 2 × 2 factorial design consists of F. graminearum contamination at 1.0104 cfu/g (UCT (no contamination) vs. CT (contamination)) and inoculant application at 2.1 × 105 cfu/g (CON (no inoculant) vs. INO (inoculant)) used in this study. After 90 days of ensiling, the contents of CP, NDF, and ADF increased (p<0.05) by F. graminearum contamination, while IVDMD, acetate, and aerobic stability decreased (p<0.05). Meanwhile, aerobic stability decreased (p<0.05) by inoculant application. There were interaction effects (p<0.05) on IVNDFD, NH3-N, LAB, and yeast, which were highest in UCT-INO, UCT-CON, CT-INO, and CT-CON & INO, respectively. In conclusion, this study found that mold contamination could negatively impact silage quality, but isolated inoculants had limited effects on IVNDFD and yeast.

Compared to operational wastes, nuclear power plant (NPP) decommissioning wastes are generated in larger quantities within a short time and include diverse types with a wider range of radiation characteristics. Currently used 200 L drums and IP-2 type transport containers are inefficient and restrictive in packaging and transporting decommissioning wastes. Therefore, new packaging and transport containers with greater size, loading weight, and shielding performance have been developed. When transporting radioactive materials, radiological safety should be assessed by reflecting parameters such as the type and quantity of the package, transport route, and transport environment. Thus far, safety evaluations of radioactive waste transport have mainly targeted operational wastes, that have less radioactivity and a smaller amount per transport than decommissioning wastes. Therefore, in this study, the possible radiation effects during the transport from NPP to disposal facilities were evaluated to reflect the characteristics of the newly developed containers and decommissioning wastes. According to the evaluation results, the exposure dose to transport workers, handling workers, and the public was lower than the domestic regulatory limit. In addition, all exposure dose results were confirmed, through sensitivity analysis, to satisfy the evaluation criteria even under circumstances when radioactive materials were released 100% from the container.

As nuclear power plants are operated in Korea, low and intermediate-level radioactive wastes and spent nuclear fuels are continuously generated. Due to the increase in the amount of radioactive waste generated, the demand for transportation of radioactive wastes in Korea is increasing. This can have radiological effect for public and worker, risk assessment for radioactive waste transportation should be preceded. Especially, if the radionuclides release in the ocean because of ship sinking accident, it can cause internal exposure by ingestion of aquatic foods. Thus, it is necessary to analyze process of internal exposure due to ingestion. The object of this study is to analyze internal exposure by ingestion of aquatic foods. In this study, we analyzed the process and the evaluation methodology of internal exposure caused by aquatic foods ingestion in MARINRAD, a risk assessment code for marine transport sinking accidents developed by the Sandia National Laboratory (SNL). To calculate the ingestion internal exposure dose, the ingestion concentrations of radionuclides caused by the food chain are calculated first. For this purpose, MARINRAD divide the food chain into three stages; prey, primary predator, and secondary predator. Marine species in each food chain are not specific but general to accommodate a wide variety of global consumer groups. The ingestion concentrations of radionuclides are expressed as an ingestion concentration factors. In the case of prey, the ingestion concentration factors apply the value derived from biological experiments. The predator's ingestion concentration factors are calculated by considering factors such as fraction of nuclide absorbed in gut, ingestion rate, etc. When calculating the ingestion internal exposure dose, the previously calculated ingestion concentration factor, consumption of aquatic food, and dose conversion factor for ingestion are considered. MARINRAD assume that humans consume all marine species presented in the food chain. Marine species consumption is assumed approximate and conservative values for generality. In the internal exposure evaluation by aquatic foods ingestion in this study, the ingestion concetration factor considering the food chain, the fraction of nuclide absorbed in predator’s gut, ingestion rate of predator, etc. were considered as influencing factors. In order to evaluate the risk of maritime transportation reflecting domestic characteristics, factors such as domestic food chains and ingestion rate should be considered. The result of this study can be used as basis for risk assessment for maritime transportation in Korea.

Kori unit 1, the first PWR (Pressurized Water Reactor) in Korea, was permanent shut down in 2017. In Korea, according to the Nuclear Safety Act, the FDP (Final Decommissioning Plan) must be submitted within 5 years of permanent shutdown. According to NSSC Notice, the types, volumes, and radioactivity of solid radioactive wastes should be included in FDP chapter 9, Radioactive Waste Management, Therefore, in this study, the types depending on generation characteristics and radiological characterization methods and process of solid radioactive waste were analyzed. Solid radioactive waste depending on the characteristics of the generation was classified into reactor vessel and reactor vessel internal, large components, small metals, spent nuclear fuel storage racks, insulation, wires, concrete debris, scattering concrete, asbestos, mixed waste, soil, spent resins and filters, and dry active waste. Radiological characterization of solid radioactive waste is performed to determine the characteristics of radioactive contamination, including the type and concentration of radionuclides. It is necessary to ensure the representativeness of the sample for the structures, systems and components to be evaluated and to apply appropriate evaluation methods and procedures according to the structure, material and type of contamination. Therefore, the radiological characterization is divided into concrete and structures, systems and components, and reactor vessel, reactor vessel internal and bioshield concrete. In this study, the types depending on generation characteristics and radiological characterization methods and process of solid radioactive waste were analyzed. The results of this study can be used as a basis for the preparation of the FDP for the Kori unit 1.

Wolsong unit 1, the first PHWR (Pressurized Heavy Water Reactor) in Korea, was permanent shut down in 2019. In Korea, according to the Nuclear Safety Act, the FDP (Final Decommissioning Plan) must be submitted within 5 years of permanent shutdown. According to NSSC Notice, the types, volumes, and radioactivity of solid radioactive wastes should be included in FDP chapter 9, Radioactive Waste Management, Therefore, in this study, activation assessment and waste classification of the End shield, which is a major activation component, were conducted. MCNP and ORIGEN-S computer codes were used for the activation assessment of the End shield. Radioactive waste levels were classified according to the cooling period of 0 to 20 years in consideration of the actual start of decommissioning. The End shield consists of Lattice tube, Shielding ball, Sleeve insert, Calandria tube shielding sleeve, and Embedment Ring. Among the components composed for each fuel channel, the neutron flux was calculated for the components whose level was not predicted by preliminary activation assessment, by dividing them into three channel regions: central channel, inter channel, and outer channel. In the case of the shielding ball, the neutron flux was calculated in the area up to 10 cm close to the core and other parts to check the decrease in neutron flux with the distance from the core. The neutron flux calculations showed that the highest neutron flux was calculated at the Sleeve insert, the component closest to the fuel channel. It was found that the neutron flux decreased by about 1/10 to 1/20 as the distance from the core increased by 20 cm. The outer channel was found to have about 30% of the neutron flux of the center channel. It was found that no change in radioactive waste level due to decay occurred during the 0 to 20 years cooling period. In this study, activation assessment and waste classification of End Shield in Wolsong unit 1 was conducted. The results of this study can be used as a basis for the preparation of the FDP for the Wolsong unit 1.

In Korea, the construction of dry storage facilities for spent nuclear fuel is being promoted through the 2nd basic plan for high-level radioactive waste management. When operating dry storage facilities, exposure dose assessment for workers should be performed, and for this, exposure scenarios based on work procedures should be derived prior. However, the dry storage method has not yet been sufficiently established in Korea, so the work procedure has not been established. Therefore, research is needed to apply it domestically based on the analysis of spent nuclear fuel management methods in major overseas leading countries. In this study, the procedure for receiving and storing spent nuclear fuel in a concrete overpack-based storage facility was analyzed. Among the various spent nuclear fuel management systems, the metal overpack-based HI-STAR 100 system and the concrete overpackbased HI-STORM 100 system are quite common methods in the United States. Therefore, in this study, work procedures were analyzed based on each final safety analysis report. First, the HI-STAR 100 overpack enters the facility and is placed in the transfer area. Remove the impact limiter of the overpack and install the alignment device on the top of the overpack. Place the HI-TRAC, an on-site transfer device, on top of the alignment unit and remove the lids of the two devices to insert the canister into the HI-TRAC. When the canister transfer is complete, reseat the lid to seal it, and disconnect the HI-TRAC from the HI-STAR 100. Raise the canister-loaded HI-TRAC over the alignment device on the top of the HI-STORM 100 overpack and remove the lids of the two devices that are in contact. Insert the canister into the HI-STORM 100 and reseat the lid. The HI-STORM 100 loaded with spent nuclear fuel is transferred to the designated storage area. In this study, the procedure for receiving and storing spent nuclear fuel in a concrete overpack-based storage facility was analyzed. The main procedure was the transfer of canisters between overpacks, and it was confirmed that HI-TRAC was used in the work procedure. The results of this study can be used as basic data for evaluating the exposure dose of operating workers for the construction of dry storage facilities in Korea.