Lumpy Skin Disease (LSD) and Foot-and-Mouth Disease (FMD) cause substantial economic losses on the livestock industry. Therefore, vaccinations have been implemented as the control strategy in endemic countries. However, the potential adverse effects of administering vaccines for both diseases simultaneously have not been thoroughly evaluated. The aim of this study was to assess the impact of vaccinating dairy cows with either or both LSD and FMD vaccines on milk production and physiological parameters such as milk temperature, rumination time and body weight. The experimental groups were divided into four according to the injection materials: 1) saline, 2) LSD vaccine, 3) FMD vaccine, and 4) both vaccines. The impact of vaccination on milk yield and physiological parameters was evaluated daily until 12 days post-vaccination, and milk components were analyzed twice, once per week. Among the experimental groups as well as each vaccine group, no statistically significant differences (p < 0.05) were observed at milk yield, milk components, or milk temperature. This suggests that simultaneous vaccination of LSD and FMD can be administered without adverse effects.

진저 비어는 생강과 설탕을 이용한 발효 음료로, 가볍 게 탄산화된 특유의 매운맛이 특징이며, 주로 가정에서 직 접 만들어진다. Ginger bug라는 스타터 컬처를 사용한 자 연적 발효 과정을 통해 만들어지며, 이는 상업용 음료와 달리 발효된 상태의 살아있는 미생물을 포함한다. 이 연 구는 두 가지 다른 방법으로 가정용 진저 비어를 직접 제 조하여 진저 비어의 미생물 군집의 변화를 분석하고자 하 였다. 레시피 1과 2의 발효 결과, 총 균수(aerobic plate count, APC)는 최대 6 log CFU/mL에 도달했고, 효모와 곰팡이 수(yeast and mold, YM)는 6.5 log CFU/mL로 가 장 높았다. 레시피 2에서는 진저 비어를 만들기 전에 ginger bug를 발효하였으므로 알코올 함량이 0.655%까지 증가한 반면, 레시피 1에서는 0.15% 미만이었다. 다양성 분석 결 과, ginger bug에서 높은 수준의 Enterobacteriaceae가 발견 되어 발효 과정과 재료 취급이 미생물 군집 변화에 영향 을 미쳤음을 시사했다. 생강과 진저 비어 전반에서 Lactococcus가 낮은 수준으로 검출되었고, 진저 비어에서 는 셀룰로오스를 분해하는 Trabulsiella 균주가 발견되어 프로바이오틱스 가능성을 시사하였다. 본 연구는 진저 비 어의 미생물 군집에 대한 최초의 연구로, 진저 비어 제조 시 재료로부터 기원한 미생물이 어떻게 변화하는지에 대 한 통찰을 제공한다. 또한, 다양한 환경에서의 발효 조건 이 미생물 군집과 제품의 품질에 미치는 영향을 탐구하는 데 기여할 것이다. 연구 결과는 진저 비어의 품질 향상에 대한 향후 연구에 중요한 자료를 제공할 것이다.

Maturity at harvest is the key factor influencing storage life and the final quality of fruit. This study examined how the firmness of ‘Sweet Gold’ kiwifruit at harvest affects its post-ripening characteristics to create a maturity chart. Throughout the storage period, firmness decreased in all categories of fruit: hard, medium, and soft. ‘Soft’ fruits lost 40% of their firmness within 2 days after harvest, whereas ‘hard’ fruits remained firmer than soft fruits throughout the storage period and had the least soluble solids content, indicating a slower ripening progression. The acidity of ‘soft’ kiwifruit was very low from the day of harvest, suggesting that it was utilized as a respiratory substrate during ripening. The a-values (indicating redness) for ‘soft’ fruits gradually increased until day 6, stabilizing thereafter. ‘Soft’ fruits exhibited the highest ethylene production rate throughout storage. They showed a climacteric rise in ethylene on day 13, compared to ‘medium’ and ‘hard’ fruits, which exhibited increases on days 19 and 21, respectively. This data can help determine the optimal ethylene treatment duration for ripening ‘Sweet Gold’ kiwifruit. The firmness of ‘Sweet Gold’ kiwifruit at harvest is a crucial factor in determining its marketability and can effectively serve as a maturity index to estimate its shelf life.

After the permanent shut down of Kori Unit 1, various decommissioning activities will be implemented, including decontamination, segmentation, waste management, and site restoration. During the decommissioning period, waste management is among the most important activities to ensure that the process proceeds smoothly and within the expected timeframe. Furthermore, the radioactive waste generated during the operation should be sent to a disposal facility to complete the decommissioning project. Square and cylindrical concrete re-package drums were generated during the 1980s and 1990s. The square, containing boron concentrates, and cylindrical, containing spent resin, concrete re-package drums have been stored in a radioactive waste storage building. Homogeneous radioactive waste, including boron concentrates, spent resin, and sludge, should be solidified or packaged in high-integrity containers (HICs). This study investigates the sequential segmentation process for the separation of contaminated and non-contaminated regions, the re-packaging process of segmented or crushed cement-solidified boron concentrate, and re-packaging in HICs. The conceptual design evaluates the re-packaging plan for the segmented and crushed cement-solidified waste using HICs, which is acceptable in a disposal facility, and the quantity of generated HICs from the treatment process.

본 논문에서는 몬스터에게 붙잡히지 않고 키를 찾아서 연구소를 탈출하는 공포 방탈출 게임을 제안한다. 제 안하는 게임을 진행한30명의 게임 플레이 로그 데이터 분석 결과와 설문조사 결과를 바탕으로, 제안하는 게 임의 특징을 분석한 결과는 다음과 같다. 첫째, 제안하는 게임은 다양한 아이템, 액션, 탈출 경로를 제공한다. 제안하는 게임이 숨을 곳도 많고 다양한 상호작용을 제공한다고 설문에서4점 이상 주었다. 또한, map의 footprint를 분석한 결과, 플레이어는 다양한 경로를 통해 키를 찾아서 탈출하였다. 둘째, 제안하는 게임은 외 관으로 기능을 추론할 수 있는 직관적인 오브젝트를 제공한다. 따라서, 플레이어는 시각적 공간 및 게임 아 이템 용도를 쉽게 파악하여 조작할 수 있다. 설문조사 결과에서, 플레이어는 조작감 관련 항목의 점수를4점 이상을 주었다. 셋째, 제안하는 게임에서 플레이어는 아이템이 충분할 때보다는 부족할 때 더 몰입을 잘 한 다. 게임 플레이 로그 데이터 분석 결과와 설문조사 결과에 따르면, 플레이어는 아이템이 부족할 때 더 크게 공포를 느끼고 상황에 몰입하여, 더 적극적으로 행동하게 되고 더 민감하게 반응한다.

Alpha activities can be used for categorization, transportation, and disposal of radioactive waste generated from the operation of nuclear facilities including nuclear power plants. In order to transport and dispose of such low- and intermediate-level radioactive waste (LILW) to the Wolsong LILW Disposal Center (WLDC) at Gyeongju, the gross alpha concentration of an individual drum should be determined according to the acceptance criteria. In addition, when the gross alpha concentration exceeds 10 Bq/g, the inventory of the comprising alpha emitters in the waste is to be identified. Gross alpha measurements using a proportional counter are usually straightforward, inexpensive, and high-throughput, so they are broadly used to assay the total alpha activity for environmental, health physics, and emergency-response assessments. However, several factors are thoughtfully considered to obtain a reliable approximate for the entire alpha emitters in a sample, which include the alpha particle energy of a particular radionuclide, the radionuclide that is used as a calibration standard, the uniformity of film in a planchet, time between sample collection and sample preparation, and time between sample preparation and counting. Korea Atomic Energy Research Institute (KAERI) have evaluated the inventory of radionuclides in low-level radioactive waste drums to send every year hundreds of them to the WLDC. In this presentation, we revisit the gross alpha measurement results of the drums transported to WLDC in the past few years and compare them with the concentrations of alpha emitters measured from alpha spectrometry and gamma spectrometry. This study offers an insight into the gross alpha measurement for radioactive waste regarding calibration source, self-absorption effect, composition of alpha emitters, etc.

Spent ion exchange resins have been generated during the operation of nuclear facilities. These resins include radioactive nuclides. It is needed to fabricate them into a stable form for final disposal. Cement solidification process is a useful method for the fabrication of them into a waste form for final disposal. In this study, proper conditions for the fabrication of them into a stable waste form were determined using the cement solidification process. In-drum waste forms were then produced at the conditions, where the stability of representative samples was evaluated for final disposal. The samples were satisfied to the Waste Acceptance Criteria for low and intermediate level radioactive waste disposal sites. This result can be utilized to derive optimal conditions for the fabrication of spent ion exchange resins into a final disposal form.

Ring Tensile Test (RTT) is mainly performed for comparing tensile strength and total strain between nuclear fuel cladding specimens under various initial conditions. Through RTT, the loaddisplacement (F-D) curve obtained from the uniaxial tensile test can also be obtained. However, the Young’s modulus estimated from the gradient of the straight portion is much lower than general value of materials. The reasons include tensile machine compliance, slack in the fixtures, or elastic deformation of the fixtures and the tooling. Another reason is that the bending of the test part in the ring is stretched with two pieces of tools. Although the absolute value of the Young’s modulus is smaller than the actual value, it is applicable to calculate the ratio of the Young’s moduli of different materials, that is, the relative value. The Young’s modulus, or slope of the linear section, varies slightly depending on which location data is used and how much data is included. In order to obtain a more accurate ratio of Young’s moduli between materials using the RTT results, a post-processing method for the ring tensile test results that can prevent such human errors is proposed as follows. First, the slope of the linear section is obtained using the displacement and load when the load increase is the largest and the displacement and load of the position that is 95% of the maximum load increase. To replace the section where the ring-shaped specimen is stretched at the beginning of the F-D curve, a straight line equal to the slope of the linear section is drawn to the displacement axis from the position of maximum load increase and moved to the origin to obtain the final F-D curve for a RTT. Lastly, the yield stress uses the stress at the point where the 0.2% offset straight line and the F-D curve meet as suggested in the ASTM E8/E8M-11 “Standard test methods for tensile testing of metallic materials”. RTT results post-processing method was coded using FORTRAN language so that it could be performed automatically. In addition, sensitivity analysis of the included data range on the Young’s modulus was performed by using the included data range as 90%, 85%, and 80% of the maximum load increase.

In nuclear fuel development research, consideration of the back-end cycle is essential. In particular, a review of an in-reactor performance of nuclear fuel related to the various degradation phenomena that can occur during spent fuel dry storage is an important area. The important factors affecting the degradation of zirconium-based cladding during dry storage are the cladding’s hydrogen concentration and rod internal pressure after irradiation. In this study, a preliminary analysis of the in-reactor behavior of the HANA cladding, which has been developed and is currently undergoing licensing review, was performed, and based on this result, a comparative analysis between nuclear fuel with HANA cladding and current commercial fuel under storage conditions was performed. The results show that the rod internal pressure of nuclear fuel with HANA cladding is not significantly different from that of commercial cladding, and the hydrogen concentration in the cladding tends to reduce due to the increased corrosion resistance, so fuel integrity in a dry storage conditions is not expected to be a major problem. Although the lack of cladding creep data under dry storage conditions, the results from the Halden research reactor test comparing in-reactor creep behavior with Zircaloy-4 showed that there is sufficient margin for degradation due to creep during storage.

Many countries have used nuclear power to generate electricity. Uranium-235, which is used as fuel in nuclear power plants, produces many fission products. Among them, iodine-129 is problematic due to its long half-life (1.57×107 years) and high diffusivity in the environment. If it is released into the environment without any treatment, it could have a major impact on humans and ecosystems. Therefore, it must be treated into a stable form through capture and solidification. Iodine can be captured in the form of AgI through silver-loaded zeolite filters in off-gas treatment processes. However, AgI could be decomposed in the reducing atmosphere of groundwater, so it must be converted into a stable form. In this study, Al2O3, Bi2O3, PbO, V2O5, MoO3, or WO3 were added to the iodine solidification matrix, AgI-Ag2O-TeO2 glass. The glass precursors were mixed to the appropriate composition and placed in an alumina crucible. After heat treatment at 800°C for 1 hour, the melt was quenched in a carbon crucible. The leaching behavior and thermal properties of the glass samples were evaluated. The PCT-A test for leaching evaluation showed that the normalized releases of all elements were below 2 g/m2, which satisfied the U.S. glass wasteform leaching regulations. Diffrential scanning calorimetry (DSC) was performed to evaluate the thermal properties of all glass samples. The addition of MoO3 or WO3 to the AgI-Ag2O-TeO2 glass increased the glass transition temperature (Tg) and crystallization temperature (Tc) while maintaining the glass stability. The similar relative electro-static filed values of MoO3, and WO3 which are approxibately three times that of the glass network former TeO2, could provide sufficient force to the TeO2 interacting with the non-bridging oxygen forming Te-O-M (M=V, Mo, W) links. The high electrostatic forces of Mo and W increased the glass network cohension and prevented the crystallization of the glass.

The decommissioning of nuclear facilities produces various types of radiologically contaminated waste. In addition, dismantlement activities, including cutting, packing, and clean-up at the facility site, result in secondary radioactive waste such as filters, resin, plastic, and clothing. Determining of the radionuclide content of this waste is an important step for the determination of a suitable management strategy including classification and disposal. In this work, we radiochemically characterized the radionuclide activities of filters used during the decommissioning of Korea Research Reactors (KRRs) 1 and 2. The results indicate that the filter samples contained mainly 3H (500–3,600 Bq·g−1), 14C (7.5–29 Bq·g−1), 55Fe (1.1– 7.1 Bq·g−1), 59Ni (0.60–1.0 Bq·g−1), 60Co (0.74–70 Bq·g−1), 63Ni (0.60–94 Bq·g−1), 90Sr (0.25–5.0 Bq·g−1), 137Cs (0.64–8.7 Bq·g−1), and 152Eu (0.19–2.9) Bq·g−1. In addition, the gross alpha radioactivity of the samples was measured to be between 0.32–1.1 Bq·g−1. The radionuclide concentrations were below the concentration limit stated in the low- and intermediatelevel waste acceptance criteria of the Nuclear Safety and Security Commission, and used for the disposal of the KRRs waste drums to a repository site.

Uranium-235, used in nuclear power generation, produces a lot of radioactive waste. Among radioactive waste nuclides, I-129 is problematic due to its long half-life (1.57×107 y) with high mobility in the environment. It should be captured and immobilized into a geological disposal environment through a stable waste form. In this study, various additives including Al, Bi, Pb, V, Mo and W were added to silver tellurite glass to prepare a matrix for immobilizing iodine, and its thermal and leaching properties were evaluated. To prepare glass, the glass precursor mixture was placed in alumina crucibles and heated at 800°C for 1 h. Except for aluminum, there was no significant loss of constituent elements. The loading of iodine in the matrix was approximately 11-15% by weigh, excluding oxygen. The normalized releases of all the elements obtained by PCT-A were below the order of 10-1 g/m2, which satisfies US regulation (2 g/m2). Differential scanning calorimetry was performed to evaluate the thermal properties of the glass samples. The glass transition temperature (Tg) increased by adding such as V2O5, MoO3, or WO3. The similar relative electrostatic field values of V2O5, MoO3, and WO3 could provide sufficient electro static field to the TeO2 interacting with the non-bridging oxygen forming Te-O-M (M = V, Mo, W) links. The addition of MoO3 or WO3 in the silver tellurite glass system increased glass transition temperature (Tg) and crystallization temperature (Tc) while maintaining the glass stability.