This study evaluated the immunogenicity of the Bacillus Calmette-Guérin (BCG) vaccine in a guinea pig model to refine preclinical assessment methods. 24 guinea pigs were divided into four groups for immunohistochemical, histopathological, and molecular analyses, including qRT-PCR and ELISA. The ELISA results revealed significant elevations in interleukin 2 (IL-2), interferon-gamma (IFN- ), and tuberculosis-specific antibodies in vaccinated guinea pigs, particularly γ notable after 6 weeks. Although lung cytokine levels remained unchanged, spleen gene expression showed significant differences in interleukin-17, interleukin-12, interleukin-1β, and C-X-C motif chemokine ligand 10 after 6 weeks. Immunohistochemistry revealed peak IL-2 expression at 8 weeks and significant IFN-γ and TNF-α expression at 6 weeks. This study confirmed the effectiveness of BCG vaccine in guinea pigs, providing crucial insights for future tuberculosis vaccine development and standardizing immune response indicators.

본 연구는 언어의 공공성과 전문성에 대한 문제의식에서 기획되었다. 이를 위해 전문 영역인 유전자변형생물체[LMO] 관련 공공기관 홈페이지 를 비교 분석하여 공공언어 사용 실태를 파악하였다. 그 결과 다음과 같 은 시사점을 이끌어 낼 수 있었다. 장르적 측면에서 LMO 홈페이지들은 ‘안내문’의 특성을 보이며, 주제어를 홈페이지 화면 전면에 제시하여 공공 언어의 수용성과 접근성을 제고하였다. 다음으로 표기의 정확성 측면에 서 전문 용어 및 법령, 관련 규정, 약어(略語) 표기의 통일성과 일관성 확보가 필요하다는 것을 확인하였다. 소통성 측면에서는 LMO와 GMO의 개념 정의 및 설명 방식에 통일성과 일관성을 확보하여 전문 용어의 이 해도를 높일 것을 제안하였다. 또한 각 기관 홈페이지가 비교적 생산 목 적에 맞게 정보를 제공하고 있지만 기관 업무 특성에 따른 접근성과 편 의성 확보의 필요성을 강조하였다. 소통성을 강화하기 위해 관계기관 협 의에서 ‘공공언어적인 관점 반영’을 안건으로 선정할 것을 제언한다.

In the design of a spent-fuel (SF) storage, the consideration of burnup credit brings the benefits in safety and economic views. According to it, various SF burnup measurement systems have been developed to estimate high fidelity burnup credit, such as FORK and SMOPY. Recently, there are a few attempts to localize the SF burnup measurement system in South Korea. For the localization of SF burnup measurement systems, it is very important to build the isotope inventory data base (DB) of various kinds of SFs. In this study, we performed DeCART2D/MASTER core follow calculations and McCARD single fuel assembly (FA) burnup analyses for Hanbit unit 3 and confirmed the characteristic of the isotope inventory over burnup. Firstly, the core follow calculations for Cycles 1~7 were performed using DeCART2D/MASTER code system. The core follow calculation is very realistic and practical because it considers the design conditions from its nuclear design report (NDR). Secondly, the Monte Carlo burnup analyses for single FAs were conducted by the McCARD Monte Carlo (MC) transport code. The McCARD code can utilize continuous energy cross section library and treat complex geometric information for particle transport simulation. Accordingly, the McCARD code can provide accurate solutions for burnup analyses without approximations, but it needs huge computing resources and time burden to perform whole-core follow calculations. Therefore, we will confirm the effectiveness of the single McCARD FA burnup analyses by comparing the DeCART2D/MASTER core follow results with the McCARD solution. From the results, the use of single FA burnup analyses for the establishment of the DBs will be justified. Various FAs, that have different 235U enrichments and loading pattern of fuel rods and burnable absorbers, were considered for the burnup analyses. In addition, the results of the sensitivity analyses for power density, initial enrichment, and cooling time will be presented.

In this study, an aerosol process was introduced to produce CaCO3. The possibility of producing CaCO3 by the aerosol process was evaluated. The characteristics of CaCO3 prepared by the aerosol process were also evaluated. In the CaCO3 prepared in this study, as the heat treatment proceeded, the calcite phase disappeared. The portlandite phase and the lime phase were formed by the heat treatment. Even if the CO2 component is removed from the calcite phase, there is a possibility that the converted CO2 component could be adsorbed into the Ca component to form a calcite phase again. Therefore, in order to remove the calcite phase, carbon components should be removed first. The lime phase was formed when CO2 was removed from the calcite phase, while the portlandite phase was formed by the introducing of H2O to the lime phase. Therefore, the order in which each phase formed could be in the order of calcite, lime, and portlandite. The reason for the simultaneous presence of the portlandite phase and the lime phase is that the hydroxyl group (OH−) introduced by H2O was not removed completely due to low temperature and/or insufficient heating time. When the sufficient temperature (900°C) and heating time (60 min) were applied, the hydroxyl group (OH−) was removed to transform into lime phase. Since the precursor contained the hydrogen component, it could be possible that the moisture (H2O) and/or the hydroxyl group (OH−) were introduced during the heat treatment process.

Uranium-235, used for nuclear power generation, has brought radioactive waste. It could be released into the environment during reprocessing or recycling of the spent nuclear fuel. Among the radioactive waste nuclides, I-129 occurs problems due to its long half-life (1.57×107 y) with high mobility in the environment. Therefore, it should be captured and immobilized into a geological disposal system through a stable waste form. One of the methods to capture iodine in the off-gas treatment process is to use silver loaded zeolite filter. It converts radioactive iodine into AgI, one of the most stable iodine forms in the solid state. However, it is difficult to directly dispose of AgI itself in an underground repository because of its aqueous dissolution under reducing condition with Fe2+. It must be immobilized in the matrix materials to prevent release of iodine as a result of chemical reaction. Among the matrix glasses, silver tellurite glass has been proposed. In this study, additives including Al, Bi, Pb, V, Mo, and W were added into the silver tellurite glass. The thermal properties of each matrix for radioactive iodine immobilization were evaluated. The glasses were prepared by the melt-quenching method at 800°C for 1 h. Differential scanning calorimetry (DSC) was performed to evaluate the thermal properties of the glass samples. From the study, the glass transition temperature (Tg) was increased by adding additives such as V2O5, MoO3, or WO3 in the silver tellurite glass. The relative electro-static field (REF) values of V2O5, MoO3, and WO3 are about three times higher than that of the glass network former, TeO2. It could provide sufficient electro-static field (EF) to the TeO2 interacting with the non-bridging oxygen forming Te-O-M (M = V, Mo, W) links. Therefore, the addition of V2O5, MoO3, or WO3 reinforced the glass network cohesion to increase the Tg of the glass. The addition of MoO3or WO3 in the silver tellurite glass increased Tg and crystallization temperature (Tc) with remaining the glass stability.

To reduce the environmental burden caused by the disposal of spent nuclear fuel and maximize the utilization of the repository facility, waste burden minimization technology is currently being developed at the Korea Atomic Energy Research Institute (KEARI). The technology includes a nuclide management process that can maximize disposal efficiency by selectively separating and collecting major nuclides in spent nuclear fuel. In addition, for efficient storage facility utilization, the short-term decay heat generated by spent nuclear fuel must be removed from the waste stream. To minimize the short-term thermal load on the repository facility, it is necessary to separate heat generating nuclides such as Cs-137 and Sr-90 from the spent fuel. In particular, Sr-90 must be separated because it generates high heat during the decay process. KAERI has developed a technology for separating Sr nuclides from Group II nuclides separated through the nuclide management process. In this study, we prepared Sr ceramic waste form, SrTiO3, by using the solid-state reaction method for long-term storage for the decay of separated Sr nuclides and evaluated the physicochemical properties of the waste form. Also, the radiological and thermal characteristics of the Sr waste form were evaluated by predicting the composition of Sr nuclides separated through the nuclide management process, and the estimation of centerline temperature was carried out using the experimental thermal data and steady state conduction equation in a long and solid cylinder type waste form. These results provided fundamental data for long-term storage and management of Sr waste.