Progressive multifocal leukoencephalopathy (PML) is a demyelinating disease by John Cunningham virus (JC virus) infection in oligodendrocytes. The radiographic and clinical features, along with the identification JC in cerebrospinal fluid polymerase chain reaction, are sufficient for the diagnosis of PML in immunodeficiency. However, it is difficult to suspect PML without the patient history of immunodeficiency. A 32-year-old man presented with headache for a month without any medical history. Based on clinical and image features, the differential diagnoses included demyelinating lesion and neoplasms. Microscopically, biopsy specimen showed multifocal demyelinating and degenerative white matter, consistent with PML. Oligodendrocytes cells with increased nuclei and plum-colored inclusions were admixed with perivascular lymphocytic and histiocytic infiltration, and loss of myelin. Atypical astrocytes had large or multiple nuclei. After brain biopsy, human immunodeficiency virus infection was confirmed by serum chemiluminescent immunoassay. It is unlikely that PML would be considered without the information of immunosuppression. Therefore, it is very important to be aware of the histological features of PML.

실감형 콘텐츠 관련 연구가 활발히 진행되고 있다. AR을 중심으로 한 실감형 콘텐츠의 발전은 스마트 폰을 이용하여 실세계에 3D 가상 콘텐츠를 추가하는 것을 넘어, 대형 건축물에 Projection Mapping을 하는 Media Facade 콘텐츠로 발전되었다. 최근 Projection Mapping을 이용한 Media Facade를 활용한 광고 콘텐츠의 수요가 증가하면서 Projection Mapping콘텐츠에 대한 교육과 시스템개발이 요구 되고 있다. Projection Mapping을 처음 배우는 교육생의 경우, 투사할 사물 및 건축물을 만드는 것부터 큰 어려움으로 작용된다. 사물 및 건축물을 구축하기 위해서는 그만한 자원과 재원이 요구되며, 한번 만들고 나면 다시 자원 을 활용하여 바꾸기 쉽지 않다. 본 논문에서는 Projection Mapping 교육을 위한 Unity 3D Engine 기반 Projection Mapping Simulation을 제안한다. 제안된 Projection Mapping Simulation는 Unity 3D Engine을 기반하였기 때문에 무료로 교육생들이 사용이 가능하며, 다양한 3D Object File를 활용하여 사물 및 건축물을 나타내고 조작할 수 있다. 또한, Projection Mapping에 쓰이는 영상을 Video 뿐 아니라, 실시간 3D Rendering으로 사용할 수도 있다. 본 논문에서 제안한 Projection Mapping Simulation는 교육생들에게 공간과 사 물 및 건축물의 이해를 돋기 위한 실험의 장이 될 것이며, Projection Mapping 콘텐츠 연구 자들에게는 여러 Projection Mapping 기법을 실험 및 연구할 수 있는 장이 되길 희망한다.

Bangladeshi medicinal plants (BMP) have a history of traditional use in treating chronic inflammatory diseases, but a BMP bark’s antioxidant and anti-inflammatory effects remain largely unexplored. This study assessed methanolic extracts’ antioxidant and anti-inflammatory properties from the bark of 15 medicinal plant species native to Bangladesh. The methanol extracts of BMP bark were evaluated for their total antioxidant activity and ability to counteract inflammation induced by lipopolysaccharide (LPS) in RAW 264.7 macrophages. Among the 15 bark extracts from BMP, Albizia odoratissima (A. odoratissima), Engelhardia spicata (E. spicata), and Shorea robusta (S. robusta) showed the highest total phenolic contents and total antioxidant capacity by effectively scavenging free radicals. In particular, these three bark extracts significantly reduced the mRNA expression of LPS-induced inflammatory cytokines and enzymes inducible by inflammation in macrophages. Also, the mRNA expression of NADPH oxidase 2 was significantly suppressed by the three bark extracts in LPS-induced RAW 264.7 macrophages. These results suggest that out of the 15 bark extracts obtained from medicinal plants in Bangladesh, the extracts from A. odoratissima, E. spicata, and S. robusta exhibit substantial total antioxidant capacity by efficiently scavenging free radicals and also inhibit LPS-induced inflammation in macrophages.

In May 2023, a free-living marine nematode species from the genus Parapinnanema was identified in the subtidal zone of Ulleungdo Island, the East Sea, Korea. Specimens were collected using the Smith-McIntyre Grab. These specimens exhibited close similarities to Parapinnanema imbricatum from the sublittoral of Moneron Island, particularly in terms of general characteristics, such as the detailed structure of the buccal cavity, the complex and ringed structure of the cuticle, the copulatory apparatus, spinneret, and the female genital system. However, the Korean specimens of Parapinnanema imbricatum also displayed distinctive features compared to the original description, including a relatively elongated body (3,317-4,339 μm vs. 3,100-4,200 μm) and a narrower body width (66-77 μm vs. 71-85 μm). Additionally, the diameter of the head was relatively shorter (24-29 μm vs. 28-36 μm). This paper offers a comprehensive morphological description, along with illustrations and DIC photomicrographs, of P. imbricatum from Korean waters.

산오이풀(Sanguisorba hakusanensis)은 한국의 자생식물 이며 정원소재로써 가치가 있지만, 생육 및 생리적 특성 및 정 원 적응 여부에 대하여 알려진 정보가 많지 않아 이용에 어려 움을 겪고 있다. 본 연구에서는 자생식물인 산오이풀의 관수 주기 및 NaCl 농도에 따른 생장, Fv/Fm, NPQ, 성분 변화, 무기성분 변화를 조사하여 내건 및 내염성 보유 여부, 생육 한 계 범위, 스트레스 환경에서 생육을 유지하기 위한 반응을 파 악하고자 했다. 실험 결과 NaCl 무처리구의 관수주기에 따른 성분 분석에서 엽록소 함량의 감소를 제외하고 유의한 차이가 나타나지 않았으나 이는 토양수분함량이 건조 스트레스를 유 발할 정도로 감소하지 않았기 때문으로 보인다. 염 처리에서 는 2주 이후 급격한 스트레스 반응이 나타났고 3주차부터 고 사하기 시작하여 6주차에 모든 개체가 최종 고사했다. 이러한 결과는 2주까지 염 스트레스에 의해 유발되는 2가지 스트레 스 중 초기에 나타나는 삼투 스트레스에는 저항하였으나 이후 나타나는 NPQ의 감소 등 이온 스트레스에 의해 유발된 광합 성 기구 붕괴로 인해 정상적인 생육을 유지할 수 없었기 때문 에 나타난 것으로 보인다. 그러나 무기이온 분석은 이온 스트 레스에 저항하기 위한 메커니즘의 존재 가능성을 시사하였다. 상대적으로 염 농도가 낮을 때에는 세포내 Ca2+ 및 K+ 수준이 높았는데, 이는 Ca2+ 수준이 높아짐에 따라 Na+를 세포 밖으 로 방출시키는 단백질, Na+를 K+와 함께 수송하는 단백질이 기능하여 Na+축적을 지연시키는 반응이 있었음을 시사한다. 그러나 NaCl을 고농도로 처리했을 때는 이러한 반응이 관찰 되지 않았다. 따라서 산오이풀은 염 스트레스에 의해 야기되 는 삼투 스트레스에 강한 저항성을 가지고 있고 이온 독성을 줄이기 위한 메커니즘으로 Na+ 세포내 축적을 지연시키는 것으로 보이지만, 심한 염 스트레스를 받았을 때 나타나는 급격 한 반응에서 이러한 메커니즘이 기능하지 못하고 이온독성에 매우 취약한 것으로 여겨진다. 본 연구를 통해 자생식물인 산 오이풀의 활용을 늘리는 데 기초적인 자료를 제공할 수 있을 것으로 생각된다.

식물의 흡수를 통한 공기오염물질 제거는 생육 상태에 따라 그 효과가 달라진다. 실내에서 토양수분의 공급은 식물의 생 육을 위한 기본적인 관리 사항이다. 따라서 본 연구는 토양수 분함량에 따른 생리적 반응이 가스상 공기오염물질인 톨루엔 저감에 미치는 영향을 구명하고, 최적의 생육과 공기 정화 효 과를 위한 적정 토양수분함량을 찾고자 수행하였다. 이를 위 해 스파티필름과 파키라를 사용하여 40일 동안의 평균 토양 수분함량을 25%, 20%, 15%, 10%로 처리한 후 양자수율, 광 합성률, 기공전도도, 증산량 등 생리적 지수와 엽면적당 톨루 엔 저감량을 측정하였다. 그 결과 스파티필름은 토양수분함량 을 20~25%로 관리할 때 생육이 양호하고 최적의 톨루엔 저 감 효과를 얻을 수 있을 것으로 판단되며, 10% 이하 건조에 대한 주의가 요구된다. 반면 파키라는 토양수분함량 20% 이 하 처리구에서 톨루엔 저감량이 증가하였으나 10% 처리구에 서 생장량이 저하될 가능성이 있으므로, 공기 정화와 생육을 위한 최적 토양수분함량은 15~20% 범위이며, 25% 이상으로 장기간 유지하는 것은 과습을 유발할 가능성이 있는 것으로 판단된다.

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.

This study explores the impact of metal doping on the surface structure of spent nuclear fuels (SNFs), particularly uranium dioxide (UO2). SNFs undergo significant microstructural changes during irradiation, affecting their physical and chemical properties. Certain elements, including actinides and lanthanides, can integrate into the UO2 lattice, leading to non-stoichiometry based on their oxidation state and environmental conditions. These modifications are closely linked to phenomena like corrosion and oxidation of UO2, making it essential to thoroughly characterize SNFs influenced by specific element doping for disposal or interim storage decisions. The research employs X-ray diffraction (XRD), scanning electron microscopy (SEM), and Raman spectroscopy to investigate the surface structure of UO2 samples doped with elements such as Nd3+, Gd3+, Zr4+, Th4+, and ε-particles (Mo, Ru, Pd). To manufacture these samples, UO2 powders are mixed and pelletized with the respective dopant oxide powders. The resulting pellet samples are sintered under specific conditions. The XRD analysis reveals that the lattice parameters of (U,Nd)O2, (U,Gd)O2, (U,Zr)O2, and (U,Th)O2 linearly vary with increasing doping levels, suggesting the formation of solid solutions. SEM images show that the grain size decreases with higher doping levels in (U,Gd)O2, (U,Nd)O2, and (U,Zr)O2, while the change is less pronounced in (U,Th)O2. Raman spectroscopy uncovers that U0.9Gd0.1O2-x and U0.9Nd0.1O2-x exhibit defect structures related to oxygen vacancies, induced by trivalent elements replacing U4+, distorting the UO2 lattice. In contrast, U0.9Zr0.1O2 shows no oxygen vacancy-related defects but features a distinct peak, likely indicating the formation of a ZrO8-type complex within the UO2 lattice. ε-Particle doped uranium dioxide shows minimal deviations in surface properties compared to pure UO2. This structural characterization of metal-doped and ε-particle-doped UO2 enhances our understanding of spent nuclear fuel behavior, with implications for the characterization of radioactive materials. This research provides valuable insights into how specific element doping affects the properties of SNFs, which is crucial for managing and disposing of these materials safely.

Most of the C-14 produced is in the organic form, generated as methane (14CH4), methanol (14CH3OH), formaldehyde (14CH2O), and formic acid (14CO2H2). When analyzing C-14, it is transformed into the form of 14CO2, and its concentration is determined using LSC. Typical examples include the wet oxidation method, the combustion or Pyrolysis. The wet oxidation method uses strong acids and involves repeated operations, which generates large amounts of acid waste and secondary radioactive waste. The combustion method uses high temperatures, which requires an oxygen device. Pyrolysis also requires high temperature in a vacuum and catalysts. Catalysts are expensive because they are platinum-based. To compensate for these shortcomings, a C-14 analysis method using UV irradiation was developed. In this study, 100 mL of distilled water mixed with formic acid (CO2H2), potassium persulfate (K2S2O8), and silver nitrate (AgNO3) was irradiated with a 320-390 nm UV lamp to conduct a CO2 production reaction experiment. The UV range was measured using a photometer (UV Power puck II). The beaker was made of quartz in 150 mL size with three inlets : a temperature measurement, a sample inlet, and a collection tube connector. We changed the UV lamp used from a 450 W halogen lamp to a 100 W LED, which has a lower temperature and is safer. As a result of the experiment, CO2 bubbles were generated in the collection tube, due to the UV irradiation react, which uses oxidizer and catalysts. The maximum temperature of the solution irradiated with the LED UV lamp was less than 56°C. It confirmed the rate of bubble generation changed depending on the lamp distance, the amount of sample, oxidizer, and catalyst. In an experiment to confirm the reaction caused by heat, it was found that although a reaction occurred due to heat, the reaction was significantly lower than when using a UV lamp. The reproducibility experiment was conducted three times in total under the same conditions. It showed the same pattern. In the future, we plan to select mock samples, collect 14CO2 in Carbo- Sorb, and analyze them using LSC. The results of this research will be used as a technology to recover C-14 more safely and efficiently and will also be used to expand its application to the treatment of other wastes such as waste liquid and waste resin through simulated samples.

Molten chloride salts have received considerable research attention as potential nuclear fuel and coolant candidates for molten salt reactors. However, there are several challenges, especially for structural materials due to the selective dissolution of chromium (Cr) in the molten chloride salts environment. Understanding the compatibility of uranium (U), which is used as nuclear fuel in molten salt reactors, with Cr in molten chloride salts is critical for designing the molten salt reactor structure. Therefore, in this study, the cyclic voltammetry (CV) was used to investigate the electrochemical behaviors of U and Cr. The diffusion coefficients and formal potentials were obtained. The electrochemical properties of uranium and chromium were investigated by CV in molten NaCl-MgCl2 salt at 600°C. Tungsten rods for working and counter electrode, and Ag/AgCl for reference electrode were utilized in this experiment. UCl3 made from the chemical dissolution of U rods and CrCl2 (Sigma-Aldrich, 99.99%) were used. Diffusion coefficients (D) of U and Cr were calculated by measuring reduction peak current of U3+/U and Cr2+/Cr from CV curves and using the Berzins-Delahay equation; D (U3+/U) = 3.0×10-5 cm2s-1 and D (Cr2+/Cr) = 3.3×10-5 cm2s-1. The formal potentials were also calculated by using the reduction peak potential obtained from CV results; E0’ (U3+/U) = -1.173 V and E0’ (Cr2+/Cr) = -0.321 V. The ionization tendency was investigated by comparing each reduction peak potential. The reduction peak potential Ep,c was increasing order of Ep,c (U3+/U) < Ep,c (Cr2+/Cr) < Ep,c (U4+/U3+). It can be seen that in the presence of U4+ and Cr metals, the Cr in the alloy can dissolve into Cr2+, but in the presence of U3+ and Cr metals, the Cr in the alloy does not dissolve into Cr2+. By analyzing the CV curve, diffusion coefficients and formal standard potentials were obtained. The result of comparing reduction peak potentials suggests that the nuclear fuel using U4+ should be inhibited to prevent the selective dissolution of Cr.

This program aims to build a specialized and converged educational platform for the training of students in the back-end nuclear fuel cycle and cultivate integrated human resources encompassing majors, generations, and fields. To achieve this, we have established an infrastructure for integrated education and training in the radiochemistry and back-end nuclear fuel cycle and operated specialized educational courses linked with special lectures, experimental practices, and field trips. Firstly, to construct an integrated educational and training infrastructure for the back-end nuclear fuel cycle, we formed a committee of experts from both inside and outside the institution and built an advanced radiochemistry laboratory equipped with physical and chemical analysis instruments. Through a comprehensive educational program involving theory, experiments, and discussions, we have established an integrated curriculum across adjacent majors and interdisciplinary studies. We also operate short-term education and experimental training programs (e.g., summer and winter schools for the back-end nuclear fuel cycle). Secondly, the program has connected leading researchers domestically and internationally, as well as the next generation of scholars. The program offers long-term educational opportunities and internships targeting both undergraduate and graduate students. To support this, we continuously offer expert colloquiums and individual research internships. Through regular committee meetings and workshops, we focus on nurturing the integrated talents necessary for the back-end nuclear fuel cycle. Through this program, students from various fields are being trained as competent integrated human resources capable of addressing various issues in the back-end nuclear fuel cycle. It is expected that this will enable us to supply specialized technical personnel in the back-end nuclear field in line with mid-to-long-term demands.

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.

Wolsong Unit 1, a domestic heavy water reactor nuclear power plant, was permanently shut down in December 2019. Accordingly, Wolsong Unit 1 plans to prepare a Final Decommissioning Plan (FDP), submit it to the government by 2024, receive approval for decommissioning, and begin full-scale decommissioning. One of the important tasks in the decommissioning of Wolsong Unit 1 is to determine the decommissioning strategy. It is necessary to decide on a decommissioning strategy considering various factors and variables, secure the technical background, and justify it. The selection of a decommissioning strategy is best achieved through the use of formal decisionmaking assistance techniques, such as considerations related to influencing factors. It is very important to understand the basic decommissioning strategy alternatives and whether sufficient consideration has been given to situations where only a single unit is permanently shut down in a multi-unit site like Wolsong Unit 1, while the remaining units are in normal operation. As a process for selecting a decommissioning strategy, first, all considerations that could potentially affect decommissioning presented in the KINS Decommissioning Safety Review Guidelines were synthesized, influencing factors to be used in the decision-making process were determined, and the concept was defined. In order to select the most appropriate decommissioning strategy by considering various evaluation attributes of possible decommissioning alternatives (immediate dismantling and delayed dismantling), the Wolsong Unit 1 decommissioning strategy was evaluated by reflecting the AHP decision-making technique.

Bisphenol-A, also known as BPA, is commonly used as a building block for epoxy and polycarbonate plastics. However, it has been recently identified as a major source of water pollution due to its release into the water from plastic products. BPA-based resins can also contaminate the water with high concentrations of BPA, which can enter the water bodies through production units and wastewater discharge. Photocatalysis, particularly the photo-Fenton process, is an effective method for wastewater treatment and degrading pollutants. Titanium dioxide (TiO2) is usually chosen based on its high photocatalytic properties and high performance. However, its wide band gap energy is a major issue for the photocatalytic process. This means that the catalyst can only exhibit high photocatalytic performance under UV-light irradiation and usually requires an acidic pH, which limits its use. In order to address the aforementioned issues, a visible-light photoactive photo-Fenton reaction has been successfully developed to degrade bisphenol A at natural pH using H2O2. The process was highly efficient, achieving complete degradation of phenol in just three hours of visible light irradiation with Cu-MOF. This environmentally friendly Fenton process has the advantage of occurring at natural pH levels with the presence of H2O2, providing a new perspective for efficient degradation. The photocatalyst was characterized using single X-ray diffraction (SC-XRD), powder X-ray diffraction (PXRD), Fourier Transform Infrared Spectroscopy (FTIR), and UV–vis diffuse reflectance spectroscopy (DRS).

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.