경상남도농업기술원 화훼연구소에서 2021년 화색이 연황 색이며 화심이 녹색인 미니 절화용 거베라 품종 ‘크림쿠키’ (Cream Cookie)를 육성하였다. ‘크림쿠키’ 품종은 2014년 황색 미니 ‘Sun City’를 모본으로, 백색 미니 ‘Blandy’를 부 본으로 인공교배를 실시하여 육성된 품종이다. 2021년까지 생육 및 개화 특성검정과 기호도 조사를 실시하였으며 ‘크림 쿠키’의 생육 및 개화특성을 대조품종인 ‘Sun City’와 비교하 였다. ‘크림쿠키’는 연황색(RHS 4C) 꽃잎과 녹색 화심을 가진 반겹꽃 거베라 품종이다. 화폭이 7.1cm인 작은 꽃이며, 화경 장은 58.9cm였다. 화경 직경은 상부 0.4cm, 하부 0.7cm 였 다. 외부설상화의 길이는 2.8cm이며 폭은 0.8cm였다. 개화 소요일수는 65.7일로 ‘Sun City’에 비하여 18.8일 빨랐으며, 첫 개화시 엽수는 22.4매였다. 연간 주당 절화수량은 102.3 본으로 ‘Sun City’의 82.0본에 비하여 20.3본이 많았다. 절 화수명은 17.8일로 ‘Sun City’보다 4.1일 더 길었다. ‘크림쿠 키’는 화폭이 7.1cm의 미니 품종으로 연간 주당 100본 이상 절화 생산이 가능하다. 또한 수확 후 꽃 캡 씌우기, 꽃목 보강 등 추가적인 작업을 생략할 수 있어 생산비 절감이 가능하므 로 농가보급 확대가 기대된다.

Foot-and-mouth disease (FMD), which affects cloven-hoofed animals, is economically important because of its highly contagious nature. FMD virus (FMDV), the causative agent of FMD, involves seven serotypes (O, A, Asia1, C, and SAT 1-3). Serotype Asia1 is unique to the Asian territory and is subdivided into nine genetic groups (G-I-IX) based on nucleotide variations in the VP1 sequence. Asia1 Shamir, the most representative Asia1 vaccine, is not highly protective against the Asia1/MOG/05 (G-V) lineage found in North Korea in 2007. Therefore, we investigated whether a chimeric virus strain (Asia1/MOG/Shamir), in which the VP4, VP2, and VP3 sequences of Asia1/MOG/05 were combined with the VP1 sequence of Asia1 Shamir, can simultaneously protect against both viruses. We determined the optimal viral growth conditions for the commercial utilization of this chimeric virus strain. Of the three types of cell culture media, the Cellvento medium resulted in the highest amount of antigen in the samples. The chimeric strain was proliferated in a small bioreactor to produce a test vaccine, and its immunogenicity was evaluated in pigs. The virus neutralization (VN) titer against the Asia1 Shamir virus was > 1/100 after the second immunization with the chimeric vaccine in pigs. In addition, a single dose of the test vaccine resulted in a VN titer of > 1/100 against the Asia1/MOG/05 strain. Taken together, our chimeric vaccine strain provided sufficient protection against the Asia1/MOG/05 and Asia1 Shamir viruses, suggesting its potential as a novel vaccine for both these strains.

The nuclear fuel that melted during the Fukushima nuclear accident in 2011 is still being cooled by water. In this process, contaminated water containing radioactive substances such as cesium and strontium is generated. The total amount of radioactive pollutants released by the natural environment due to the nuclear accident in Fukushima in 2011 is estimated to be 900 PBq, of which 10 to 37 PBq for cesium. Radioactive cesium (137Cs) is a potassium analog that exists in the water in the form of cations with similar daytime behavior and a small hydration radius and is recognized as a radioactive nuclide that has the greatest impact on the environment due to its long half-life (about 30 years), high solubility and diffusion coefficient, and gamma-ray emission. In this study, alginate beads were designed using Prussian blue, known as a material that selectively adsorbs cesium for removal and detection of cesium. To confirm the adsorption performance of the produced Prussian blue, immersion experiments were conducted using Cs standard solution, and MCNP simulations were performed by modeling 1L reservoir to conduct experiments using radioactive Cs in the future. An adsorption experiment was conducted with water containing standard cesium solution using alginate beads impregnated with Prussian blue. The adsorption experiment tested how much cesium of the same concentration was adsorbed over time. As a result, it was found that Prussian blue beads removed about 80% of cesium within 10-15 minutes. In addition, MCNP simulation was performed using a 1 L reservoir and a 3inch NaI detector to optimize the amount of Prussian blue. The results of comparing the efficiency according to the Prussian volume was shown. It showed that our designed system holds great promise for the cleanup and detection of radioactive cesium contaminated seawater around nuclear plants and/or after nuclear accidents. Thus, this work is expected to provide insights into the fundamental MCNP simulation based optimization of Prussian blue for cesium removal and this work based MCNP simulation will pave the way for various practical applications.

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

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

국내 유입 가능성이 높은 검역 관리해충인 Spodoptera eridania 및 S. ornithogalli는 전 세계적으로 토마토, 콩 등 여러 종의 작물을 가해하는 광식성 해충이다. 이에 따라 국내 유입 시 해당 작물에 높은 경제적 피해를 입힐 가능성 이 있으므로 신속 정확한 진단이 필요한 실정이다. 따라서 본 연구에서는 상기 두 종을 대상으로 현장 활용이 가능한 LAMP 진단법 개발을 수행하였다. 표적종 두 종 및 비표적종 11종(국내 발생 Spodoptera 종 및 동일 기주 가해종 등)의 전장유전체 정보를 확보한 후 비교 분석을 통해 각 표적종 별 특이적 영역을 확보한 후 해당 영역을 대상으로 LAMP 프라이머를 제작하였다. DNA 농도 10 ng/μL, 반응시간 40분을 기준으로 LAMP 진단을 수행한 결과, Spodoptera eridania는 5개의 LAMP 진단 마커를 개발하였고, S. ornithogalli는 3개의 LAMP 진단 마커를 개발하였다.

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.

With the recent concern regarding cellulose enhancing radionuclide mobility upon its degradation to ISA, disposal of cellulosic wastes is being held off until the disposal safety is vindicated. Thus, a rational assessment should be conducted, applying an appropriate cellulose degradation model considering the disposal environment and cellulose degradation mechanisms. In this paper cellulose degradation mechanisms and the disposal environment are studied to propose the best-suitable cellulose degradation model for the domestic 1st phase repository. For the cellulose to readily degrade, the pH should be greater than 12.5. As in the case of SKB, 1BLA is excluded from the safety assessment because the pH of 1BLA remains below 12.5. Furthermore, despite cellulose degradation occurring, it does not always produce ISA. At low Ca2+ concentration, the ISA yield rate is around 25%, but at high Ca2+ concentration, the ISA yield rate increases up to 90%. Thus, for the cellulose to be a major concern, both pH and Ca2+ concentration conditions must be satisfied. To satisfy both conditions, the cement hydration must be in 2nd phase, when the porewater pH remains around 12.5 and a significant amount of Ca2+ ion is leaching out from the cement. However, according to the safety evaluation and domestic research, 2nd phase of cement hydration for silo concrete would achieve a pH of around 12.4, dissatisfying cellulose degradation condition like in 1BLA. Thus, cellulose degradation would be unlikely to occur in the domestic 1st phase repository. To derive waste acceptance criteria, a quantitative evaluation should be conducted, conservatively assuming cellulose is degraded. To conduct a safety evaluation, an appropriate degradation model should be applied to determine the degradation rate of cellulose. According to overseas research, despite the mid-chain scission being yet to be seen in the experiments, the degradation model considering mid-chain scission is applied, resulting in an almost 100% degradation rate. The model is selected because the repositories are backfilled with cement, achieving a pH greater than 13, so extensive degradation is reasonably conservative. However, under the domestic disposal condition, where cellulose degradation is unlikely to occur, applying such model would be excessively conservative. Thus, the peeling and stopping model derived by Van Loon and Haas, which suggests 10~25% degradation rate, is reasonably conservative. Based on this model, cellulose would not be a major concern in the domestic 1st phase repository. In the future, this study could be used as fundamental data for planning waste acceptance criteria.

Concerns with colloids, dispersed 1~1,000 nm particles, in the LILW repository are being raised due to their potential to enhance radionuclide release. Due to their large surface areas, radionuclides may sorb onto mobile colloids, and drift along with the colloidal transport, instead of being sorbed onto immobile surfaces. To prevent adverse implications on the safety of the repository, the colloidal impact must be evaluated. In this paper, colloid analysis done by SKB is studied, and factors to be considered for the safety assessment of colloids are analyzed. First, the colloid generation mechanism should be analyzed. In a cementitious repository, due to a highly alkaline environment, colloid formation from wastes may be promoted by the decomposition of organic materials, dissolution of inorganic materials, and corrosion of metals. Radiolysis is excluded when radionuclide inventory is moderate, as in the case of SKB. Second, colloid stability should be evaluated to determine whether colloids remain in dispersion. Stable colloids acquire electric charges, allowing particles to continuously repel one another to prevent coagulation. Thus, stability depends on the pH and ionic condition of the surroundings, and colloid composition. For instance, under a highly alkaline cementitious environment, colloids tend to be negatively charged, repelling each other, but Ca2+ ion from cement, acting as a coagulant, makes colloid unstable, promoting sedimentation. As in the case of SKB, the colloidal impact is assumed negligible in the silo, BMA, and BTF due to their extensive cement contents, but for BLA, with relatively less cement source, the colloidal impact is a potential concern. Third, colloid mobility should be assessed to appraise radionuclide release via colloid transport. The mobility depends on the density and size of colloids, and flow velocity to commence motion. As a part of the assessment, the filtration effect should also be included, which depends on pore size and structure. As in the case of SKB, due to static hydraulic conditions and engineering barriers, acting as efficient filters, colloidal transport is expected to be unlikely. In the domestic underground repository, the highly alkaline environment would lead to colloid formation, but due to high Ca2+ concentration and low flow velocity, colloids would achieve low stability and mobility, thus colloidal impact would be a minor concern. In the future, with further detailed analysis of each factor, waste composition, and disposal condition, reliable data for safety evaluation could be generated to be used as fundamental data for planning waste acceptance criteria.

Once systems, structures and components (SSCs) of dry storage systems are classified with respect to safety function or safety significance (i.e., safety classification), appropriate engineering rules can be applied to ensure that they are designed, manufactured, maintained, managed (e.g. aging management) etc. In Unites States, the systems, structures and components (SSCs) consisting DSSs are classified into two or several grades (i.e., class A, B and C or not important to safety, and important to safety (ITS) or not important to safety (NITS)) with respect to intended safety function and safety significance. This classification methods were based on Regulatory Guide 7.10 (i.e., guidance for use in developing quality assurance programs for packaging). Also, in Korea, SSCs of DSSs should be classified into ITS and NITS in much the same as method based on Regulatory Guide 7.10. In that guidance, for providing graded approach to manage the SSCs of packaging, they were trying to classifying SSCs in accordance with radiological consequences. But there was limitations that the provided classification criteria was still qualitative, so that it was not enough for managing the SSCs according to graded approach. On the other hand, in some other nuclear facilities (i.e., nuclear power plant, radioactive waste management facility and disposal facility etc.), quantitative criteria relevant to radiological consequence (i.e., radiation doses to workers or to the public) or inventory of radioactivity are existed so that it can be applied for classifying safety classes. In summary, the study on the application safety classification that applied quantitative criteria to perform safety classification of SSCs in DSS is inadequate or insufficient. The purpose of this study is proposing the preliminary framework for estimating safety significance of SSCs in DSS which can be utilized in our further advanced studies. In this study, a framework was established to estimate the safety significance of SSCs related to radiation shielding and confinement using MCNP® 6.2 and Microsoft Excel. Referring to the methodology of IAEA Specific Safety Guide 30, we assumed severity for failures of components that could lead to degradation of the SSC’s performance. The safety class of SSC was decided based on the impact of SSC’s failure on consequences.