As the demand for nuclear power increases as a means to achieve carbon neutrality, concerns about nuclear proliferation have also grown. Consequently, significant researches have conducted to enhance nuclear non-proliferation resistance. Among these research, nuclear material attractiveness is a methodology used to evaluate how appealing a particular material is for potential use in nuclear weapons, based on the characteristics of that material. Existing nuclear material attractiveness assessments focused on materials like U, Pu, and TRU, which could be directly used in the production of nuclear weapons. However, these assessments did not consider how the properties of nuclear materials change throughout the nuclear fuel cycle, with each facility process. This study assumed a scenario of the nuclear fuel cycle of graphite reduction reactors and analyzed including enrichment facilities and PUREX. This study used the FOM (Figure-Of-Merit) method developed by LANL (Los Alamos National Laboratory) for evaluating the nuclear material attractiveness. The FOM formula consists of three parameters such as critical mass, heat content, and dose The critical mass of targe materials and the dose evaluation were conducted using the Monte Carlo N-Particle code. The heat content was calculated using the ORIGEN code embedded in the Scale code. In particular, if U-238 is dominant in the facility’s materials, such as mining and refining facilities, and critical mass evaluation is unpractical. Therefore, 1SQ (Significant Quantity) of that uranium was assumed as the critical mass value for the FOM evaluation, even though 1SQ is not identical to the critical mass As a result of this study, the attractiveness of Pu produced by PUREX among all nuclear fuel cycle facilities was 2.7616, which was the most attractive to be diverted to nuclear weapons. Through this study, it was shown that the proliferation risk of the nuclear facilities in the nuclear fuel cycle and risk of diversion among those facilities.

In this study, the process of compressing/packaging the spent filters of Kori Unit 1, which was conceptually presented in the previous study, is advanced so that disposal suitability for each step can be secure efficiently. In particular, the differences between the previous study and this study are that the disposable filters are screened using an In-Situ Object Counting System (ISOCS), and the method of collecting representative samples for development of scaling factor is specified. The process of compressing/packaging the spent filters consists of 7 stages as follows. 1) Collecting: The spent filters temporarily stored in the filter room are collected by dose and type remotely using a robot system to minimize the radiation exposure of workers according to a pre-established packaging plan. 2) Screening: The gamma activity concentration of the spent filters received by the robot system is measured by ISOCS. The spent filters below the low-level waste concentration limit and the surface dose are transferred into the compression system, while the others are returned in the filter room again. 3) Sampling: The external perforator drilling/cutting the filter was developed for sampling required for the new scaling factors. Since the sampling is collected remotely, the risk of exposure to workers can be reduced. The newly developed scaling factor will be used to verify the disposal suitability of the packages. 4) Compression: According to the pre-established plan, the spent filter collected by dose and type, is supplied to the compression system considering the dose and radionuclide inventory. Whether to additionally store the compressed filter in the drum is determined by checking the accumulated dose. 5) Immobilization: Immobilization with a safety material is necessary when inhomogeneous wastes, like spent filters, have the total radionuclide concentration with a half-life of more than 20 years is 74,000 Bq/g or more and for filling rate or non-dispersible treatment of particulates. 6) Packaging and Analysis: Waste information is labelled onto the package after the measurements of surface dose rate and surface contamination. Finally, using the drum assay system, the gamma radionuclide concentration is measured to identify at least 95% of the total radioactivity concentration of the package. 7) Temporary Storage and Delivery: The packages are moved to temporary storage in the plant prior to disposal. After establishing the plan for delivery and applying for a takeover request to KORAD, if the acceptance inspection is passed, the packages are transported to the disposal facility.

Bentonite has been considered as a buffer material in a deep geological repository for high-level radioactive waste (HLW). Bentonite may come into contacted with various chemical solutions during the long-term storage. In particular, solutions containing K+ can affect stability of bentonite (e.g., illitization). The bentonite can be gradually saturated with the inflow of groundwater, and the temperature can rise simultaneously due to the decay of HLW. This study aimed to evaluate the bentonite stability in contacted with very highly concentrated K+ solutions with different pHs at 150°C. Batch reaction tests using KJ-II bentonite were performed for 30–150 days in teflon-stainless steel reactors. De-ionized (DI) water (pH = 6.0) and 1 M KCl (pH = 6.0), and 1 M KOH (pH = 12.5) solutions were used as reaction solutions. After completing batch reaction tests, the reacted samples were analyzed using various analytical techniques. For DI water, chemical, mineralogical, and physicochemical properties of reacted samples were similar to those of unreacted samples. For 1 M KCl solutions, cation exchage for Ca by K and slight changes in mineralogical properties of reacted samples were observed, but there are no significant changes in the physicochemical properties. In contrast, for 1 M KOH solutions, changes in chemical, mineralogical, and physicochemical properties of reacted samples were observed. Results of X-ray diffraction (XRD) analysis indicated dissolution of montmorillonite and formation of zeolite minerals, which were confirmed by thermogravimetricdifferential thermal analysis (TGA-DTA) and fourier transform infrared (FTIR) analysis. These results suggest that highly concentrated K+ (1 M) solution combined with high pH (12.5) and high temperate (150°C) may affect bentonite alteration. These prelimiary experiments were intended to qualitatively evaluate the mechanism and influncing factors of the buffer material alteration under extreme experimental conditions, and it is revealed that the conditions do not reflect the actual repository environment.

Currently, treatment and disposal suitability verification methods have not been established for radioactive waste, such as spent filters temporarily stored in each plant, so the WCP (Waste Certification Program) can be applied to verify the suitability of non-conforming waste at the site. In this study, WCP components such as certification organizations, certification methods, certification documents, and quality assurance (QA) plan that should be considered when developing WCP applicable to spent filter disposal were reviewed and presented. First, a certification organization consists of a certification organization that performs certification work, a certification support organization related to waste generation and treatment, and a quality control organization for waste certification. Especially, the support organization should support the implementation of WCP, so that spent filter processing procedures such as generation information management and immobilization can be properly packaged and transported. Second, in identifying the waste characteristics of the certification method, each characteristic identification procedure and certification method of the acceptance criteria should be described, evidence examining the suitability of general, radiological, physical, chemical, and biological requirements, and processes related to measurement and sampling should be established. In identifying characteristics, satisfaction of waste form, free water requirements, and whether it is subject to immobilization should be checked priorly, and a method of confirming particulate matter and securing filling rate when packaging compressed filters should be included. It is very important to develop a technology for verifying the safety and quality of the immobilized material because immobilization of the filters can be a processing method that satisfies various characteristic criteria. Meanwhile, it is essential to collect samples and develop scaling factors to identify the nuclides of filters and prove that they are below the concentration limits. For chemical and biological requirements, the characteristics are identified through generation information documents, corrective actions are taken and documented in case of nonconformance. Third, certification documents should include immobilization procedure manual, characteristic report, and characteristic test manuals such as free water, particulate matter and filling rate, radiation measurement method manual for packages, profile, and generation documents. Fourth, the QA plan should analyze the QA system of the plants, check the QA inspection details, establish general requirements for QA of spent filter disposal, and specify step-by-step certification work QA activities. In this study, considerations to ensure the disposal suitability at all stages from generation to disposal of spent filter were presented, and development of a WCP could contribute to preventing nonconformance.

As the decommissioning of Kori Unit 1 progresses, securing technology for treatment and disposal of radioactive wastes that have not been disposed of so far, such as spent filters, is recognized as an urgent task. In this study, a method of confirming the disposal suitability of spent filters was presented by reviewing the waste characteristics as presented in the waste acceptance criteria (WAC). The waste characteristics to be satisfied to ensure disposal suitability of waste are largely classified into general requirements, solidification and immobilization requirements, radiological requirements, physical requirements, chemical requirements, and biological requirements. First, the general requirement is to prove that the prohibited waste form has not been introduced into items related to waste form and packaging, and to confirm the suitability of disposal through step-by-step packaging photos, generation information, X-ray inspection, and visual inspection. Second, in the solidification and immobilization requirements, spent filters are non-homogeneous waste, and if the total radioactivity concentration of nuclides with a half-life of more than 20 years is 74,000 Bq·g−1 or more, they must be immobilized. Third, in order to meet the characteristic criteria for nuclides and radioactivity concentration, sampling and scaling factors development are required and based on this, nuclides must be identified and demonstrated to be below the disposal concentration limits. Surface dose rate and surface contamination should be measured in accordance with standardized procedures and disposal suitability should be confirmed through document tests recording the measured values. Fourth, in order to satisfy the physical requirements of the particulate matter and filling rate characteristics, the spent filter must be immobilized, if necessary, thereby ensuring disposal suitability. Meanwhile, free water in the spent filter should be removed through pre-drying and dehydration, and the disposal suitability should be confirmed by applying a test. Fifth, the criteria for chelating agents should be checked for disposal suitability through operation records and component analysis of spent filters, and documents, that can prove harmful substances are removed in advance and no harmful substances are included in the package, should be provided. Lastly, in biological requirements, if the spent filters contain corrosive or infectious substances, they should be removed in advance and disposal suitability should be confirmed by providing documents that can prove that such substances are not included in the package.

This study established a process to ensure the disposal suitability of spent filters stored in the untreated state in Kori unit 1 and presented the following procedures and requirements for confirming the disposal suitability for each process. The process for securing spent filter disposal suitability consists of collecting spent filters, compression, immobilization, analysis and packaging, and storage stages. The requirements for confirming the acceptance criteria for each process are as follows. (1) Collecting: Since the high radioactivity spent filters are being stored in the filter room of Kori unit 1, those are collected by a remote system to minimize the exposure dose of workers due to spent filter handling. In order to satisfy the surface dose rate requirements, spent filters with a surface dose rate of 10 mSv·hr−1 or more are classified and collected, and stored temporary storage place until a separate treatment plan is determined. The checkpoints in this process are the surface dose rate, etc. (2) Compression: The collected spent filters are analyzed gamma nuclides such as Co-60 and Cs-137, using a field-applicable nuclide analyzer, and then applying the scaling factors to determine whether it is disposable. Spent filters whose radioactivity concentration is confirmed to be less than the disposal concentration limit is compressed into compression ratios determined by surface dose rate. The checkpoints in this process are nuclide information, surface dose rate, compression ratio, spent filter loading quantity, etc. (3) Immobilization: A spent filter is a non-homogeneous waste that is immobilized with a proven safety material such as cement if the total radioactivity concentration of nuclides with a half-life of more than 20 years is 74,000 Bq·g−1. Meanwhile, immobilization of inhomogeneous waste can be considered to satisfy disposal criteria such as particulate matter and filling rate. The checkpoints in this process are the immobilizing material, filling rate, etc. (4) Analysis and Packaging: Immobilized drums shall be determined to be 95% or more of the total radioactivity of waste packages by measuring the radioactivity concentration of nuclides using a nuclide analysis device. Finally, measure the surface dose rate and surface contamination of the package, and attach the package label recording the identification number, date, total radioactivity, surface dose rate, and surface contamination information to the packaging container. (5) Storage: Packaging containers are moved to and stored in a temporary waste storage or storage area before disposal.

Activated carbon fibers (ACFs) were treated by electroless plating of CuO to improve their removal performance for volatile organic compounds (VOCs). The properties of these samples(CuO@ACFs) were evaluated by X-ray photoelectron spectroscopy (XPS), BET and N2O chemisorption to determine the area and dispersion of metallic CuO. The removal efficiency for benzene was investigated by gas chromatography (GC). The breakthrough time of CuO@ACFs increased by approximately 120% compared to that of untreated ACFs at benzene of 100 ppm. CuO@ACFs removed 100% of the benzene in 20 h, indicating this material can be used as a removal technology for VOCs.

Since a cluster of pneumonia cases of unknown etiology was first reported in China in December 2019, the number of confirmed cases with coronavirus disease 2019 (COVID-19) (caused by severe acute respiratory syndrome coronavirus 2 [SARS-CoV-2]) worldwide has been increasing, and nearly two million patients are expected to have died by February 2021. Globally, COVID-19 is being considered a primary pulmonary disease, but it is more than a lung infection. This is because patients infected with SARS-CoV-2 present with a variety of clinical signs and symptoms involving many organs, ranging from fever, inflammation, myocardial injury, shock, and the development of coronary artery aneurysm. SARS-CoV-2 shows unique characteristics in its transmission, mortality, and stability in different environmental conditions. During COVID-19 pandemic, about 10 drugs have been clinically tried with none proven to be effective. With no prospect for effective drugs in the near future, there are currently possibilities of success with vaccines and convalescent plasma. Along with the prospect of vaccines and other therapeutic drugs, special precautions (isolation, testing, treating, and tracing) are strictly recommended until we return to normal situations. In this review we comprehensively analyzed the clinical presentations, characteristics, and outcomes of patients with COVID-19, and tried to assess the clinical effects of some drugs as part of the ongoing efforts to understand COVID-19 pandemic.

To enhance mechanical properties through improvement of dispersion stability of carbon black (CB) in epoxy resins, fluorine functional groups were introduced on the CB surface by fluorination. The changes in the chemical properties and dispersion stabilities after fluorination were evaluated with different partial pressures of fluorine gas. The mechanical properties of the fluorinated CB/epoxy composites were evaluated by the test of tensile, impact strengths and creep behavior. The fluorinated CB/epoxy composites showed approximately 1.6 and 1.1 times enhancement in the tensile and impact strengths compared to that of neat epoxy, respectively. Moreover, when a constant load was applied at 323 K, the fluorinated CB/epoxy composites lasted longer and had smaller strain changes than those of the raw CB/epoxy composites. Thus, well-dispersed CB by fluorination in epoxy resins effectively transfers mechanical stress.

Persistent left cranial vena cava (PLCVC) is a remnant vessel connected with the coronary sinus and draining into the right atrium. A 3-month-old intact male Bichon Frise was evaluated for the presence of a mechanical murmur auscultation in the local animal hospital. No significant clinical signs were present on physical examination except mechanical murmur. Patent ductus arteriosus (PDA) was diagnosed in the imaging procedure. During the left thoracotomy, PLCVC was found. The vascular malformation made the surgical process difficult by hiding PDA from the left thoracotomy surgical view. PLCVC and the vagus nerve was carefully dissected and lifted to secure a clear surgical view of PDA. The ductus arteriosus was ligated. Computed tomography angiography (CTA) was performed postoperatively. On CTA, left brachiocephalic vein retaining connection with the coronary sinus draining into the right atrium was observed. CTA is highly recommended for dogs with PDA to provide better postoperative results.

The baculovirus-insect cell expression system has been widely used method for the recombinant protein expression. The present study has several limitation. In this study, we constructed vectors consisting of transcriptional enhanced factor and promoter that improve the expression level. To confirm the usefulness of these vector system, Human papillomavirus (HPV) VLPs have been expressed by baculovirus hyper expression system. HPV VLPs were purified using a CaptoTM Core 700 (GE Healthcare Life Sciences) chromatography approach. Baculovirus hyper expression system production efficiency was influenced by the HPV VLPs production. HPV VLPs vaccination to BALB/c mice induced the generation of antibody confirmed by ELISA. This study could provide improvements on the vaccine production for the development of VLP vaccines high expression of useful heterologous proteins.

PURPOSES : Thermal cracking (also called low-temperature cracking) is a serious stress for asphalt pavement, especially in eastern South Korea, the northern USA, and Canada. Thermal cracking occurs when the level of thermal stress exceeds the corresponding level of low temperature strength of the given asphalt materials. Therefore, computation of thermal stress is a key factor for understanding, quantifying, and evaluating the level of low-temperature cracking resistance of asphalt pavement. In this paper, two different approaches for computing thermal stress on asphalt binder were introduced: Hopkins and Hamming’s algorithm (1967) and the application of a simple power-law function. All the computed results were compared visually; then the findings and recommendations were discussed. METHODS: Thermal stress of the tested asphalt binder was computed based on the methodology introduced in previous literatures related to viscoelastic theory. To perform the numerical analysis, MATLABTM 2D matrix-correlation and Microsoft Excel visual basic code were developed and used for the function fitting and value-minimization processes, respectively. RESULTS : Different results from thermal stress were observed with application of different computation approaches. This variation of the data trends could be recognized not only visually but also statistically. CONCLUSIONS: It can be concluded that these two different computation approaches can successfully provide upper and lower limits (i.e. boundaries) for thermal stress prediction of a given asphalt binder. Based on these findings, more reliable and reasonable thermal stress results could be provided and finally, better pavement performance predictions could also be expected.

우드펠렛은 가정용, 상업용, 산업용 등으로 사용되며 특히 화력발전소에서 석탄 대용으로 수입량이 급증하여 연간 약 2,400만톤 가량 수입되고 있는 상황이다. 우드펠렛 검역훈증제로 기존에는 메틸브로마이드(이하 MB)가 최근까지 사용되어져 왔으나, MB의 높은 흡착률과 우드펠렛의 과도한 수용비로 인해 최종가스농도 기준에 미치지 못하여 재투약하는 사례가 빈번히 발생해 문제가 되어져왔다. 이를 해결하기 위해 농림축산검역본부에서는 올해 초 MB와 포스핀을 병행으로 처리하는 방식의 소독처리기준을 신설하였다. 새로운 소독처리기준이 설정됨에 따라 배기시 안전한 배기시간 수준을 설정하기 위해 병행처리 후 수착과 탈착되는 가스농도를 조사하였다. 실내 훈증상에 우드펠렛을 채우고 약제 처리한 결과 각 훈증제의 TLV(Threshold Limit Value)-TWA(Time Weighted Average) 기준이하로 농도가 감소되는 배기시간은 25℃일 때 MB는 120시간, 포스핀은 6시간으로 측정되었고 15℃일 때 MB는 186시간, 포스핀은 19시간으로 측정되었다.

고준위방사성폐기물 심지층 처분 대상 암종으로 고려되는 화강암에서 방사성핵종의 장기 거동특성을 이해하기 위한 연구의 일환으로 KURT (KAERI Underground Research Tunnel) 화강암에 존재하는 우라늄의 용출특성에 대한 연구를 수행하였다. 반응 시작 후부터 10일 동안의 반응기간 중 다른 반응용액에 비해 CO3 2- 농도가 높은 UD-CO3 및 UD-Bg 반응용액에서 우라늄의 용출량이 다소 급격하게 증가하였다. 또한 Na 또는 Ca가 다량 함유된 반응용액에서 반응 60일 이후 우라늄 용출량이 다소 급격히 증가하였다. 각 반응용액에 의한 반응 270일까지의 우라늄의 용출량은 UD-CO3 (44.61 μg·L-1), UD-Bg (41.01 μg·L-1), UD-Na (26.87 μg·L-1), UD-Ca (20.26 μg·L-1), UD-CaSi (17.03 μg·L-1), UD-Si (10.47 μg·L-1)으로 지속적으 로 증가 하였으나, 반응 270일 이후 우라늄 용출량은 점차 감소하는 경향을 나타낸다. 이는 화강암 시료 내에 존재하는 우라늄이 반응용액과 상호반응에 의해 최대 용출될 수 있는 한계에 도달하였기 때문으로 판단된다. 우라늄 용출은 혼합된 반응 용액 내의 CO3 2- 존재와 수질의 지화학적 유형에 따라 우라늄의 용출 농도 및 용출 최대치가 나타나는 시점이 다르게 확인되 었다. 이는 시료와 반응용액의 상호반응 과정에서 용존이온의 영향에 의해 화강암시료와 반응용액 사이에 반응속도의 차이가 발생하는 것으로 판단된다.

The ride quality (i.e. smoothness) is a key factor for evaluating the construction quality of expressway asphalt pavement. Conventionally, three paving devices are widely used to control the surface layer thickness: leveling sensor (i.e. LS), short-range-surfacing-contact-ski (i.e. SSCS) and long-range-surfacing-contact-ski (i.e. LSCS). However, each of these levelling tools presents one major drawback. In the case of LS, if the original sub-layer evenness is poor, the final asphalt pavement surface and its smoothness will be negatively affected. The SSCS cannot assure satisfactory smoothness when relatively long paving section (in the order of 10 km) are paved. While the LSCS would reduce the drawback of the SSCS, its weight on the one hand and its length on the other discourage its use in the paving site especially for curved sections. In this paper, a next generation pavement smoothness leveling equipment, known as non-contact-digital-ski (i.e. NCDS) was implemented, evaluated and compared to the conventional equipment leveling device. The international Roughness Index (IRI m/km) was measured on sections paved with and without NCDS and the results visually and statistically compared. In addition, for the same sections, the modulus of the pavement layers was computed and compared by means of Falling Weight Deflectometer (i.e. FWD). It was observed that when NCDS is used for asphalt pavement overlay of existing concrete pavement, significant improvement in IRI (i.e. IRI<1.0m/km) and consistently uniform elastic modulus could be achieved compared to the conventional levelling and paving method.