전보에서 외국산에 의해 기록된 한반도산 날개응애류 26종과 필자가 채집한 미기록종 29종 도합 65종의 목록을 발표한 바 있다. 그 후의 채집품을 조사결과 다수의 미기록종이 발견되었기에 전보를 증보하여 발표한다. 이로서 한반도산 날개응애류는 29상과, 59과, 99속, 158종이 되었다. 이밖에 미정리분중에도 약 50중의 미기록종이 포함되어 있으므로 한국산 날개응애류는 200여종이 되는 셈이지만 한국에 실재하고 있는 종수는 그 배가 될 것으로 추측된다. 또한 분포를 논할 단계는 아니지만 일본과의 공통종은 69종, 구주와의 공통종은 51종, 3자공통종은 55종이 되어 한반도 고유종은 34종이 된다. 그러나 우리나라와 인접해 있는 중국대륙의 날개응애상이 밝혀지면 고유종은 줄어들고 중국과의 공통종이 많아질 것이다.

Cochylidae(가는잎말이나방과 : 신칭)은 잎말이나방상과(Tortricoidea)에 속하며 우레 나라에서는 오늘날까지 알려지지 않았던 소형의 나방 그룹이다. 그러나 이미 1886년에 영국 곤충학자 J. H. Leech의 한반도 채집시에 2종(종)이 채집되어 외국 문헌상에는 기록되어 오고 있다. 이과에 슥하는 종들은 대개가 작물의 꽃, 신소, 줄기 등 비교적 연약한 조직의 내부를 갖아 작해를 주며 전세계적으로 분포하는 작물의 주요 해충들이다. 필자는 현재까지 채집된 재료 중에서 금번에 분류 등정된 한국 미기록 10종을 발표한다.

The swelling capacity of bentonite buffers is vital in high-level radioactive waste (HLW) repositories, as it minimizes groundwater infiltration, prevents nuclides from reaching the biosphere, and stabilizes the HLW canisters. As swelling capacity is a function of temperature, understanding bentonite’s behavior at approximately 100°C (its presumed upper limit) is essential. However, research on this subject has been scarce. Hence, this study explored the effects of thermal treatment of Ca-bentonite at 105°C under injected water pressures. The results suggest a 19% reduction in “swell index” and a 35%–36% decrease in the total pressure in thermally treated bentonite. The heated samples demonstrated higher hydraulic conductivity than the non-heated ones, indicating potential performance deterioration in controlling the fluid movement. Furthermore, the injected water pressure (base pressure) was not fully transmitted to the sample owing to the difference between the base and back pressures, leading to variations in the total pressure despite maintaining a constant differential pressure. Thus, the results demonstrated a degradation in bentonite’s swelling capacity and its compromised role in safe HLW disposal, when subjected to treatment at 105°C. The insights from this research can assist in HLW repository design, while highlighting the need for further research into bentonite’s performance.

본 논문은 한국의 대표적인 오페라 작곡가 이건용의 작품에서 나타나는 한국어의 음악적 표현 을 ≪동승≫(2004)과 ≪왕자와 크리스마스≫(2010)의 몇몇 아리아 및 합창을 중심으로 분석했다. 이를 행함에 있어 한국어 고유의 특징인 장단, 고저, 억양, 음절언어적 특성, 말토막 개념 및 음악 에서의 박자 악센트와 리듬 악센트 등을 활용했다. 그 결과 이건용의 오페라에서는 문장 및 말토막이 본래의 언어적인 호흡을 유지한 채 음악화 되었으며, 한국어의 장단구조, 수식구조, 단어의 높낮이, 문장의 억양 등이 다양한 박자 악센트 및 리듬 악센트와 결합하여 해당 텍스트의 음운론적인 측면을 강화함을 알 수 있었다. 또한 이건용의 오페라는 단어 및 문장에 새로운 장단이나 악센트 등을 부여하기도 하였다. 이 경우 해당 선율들은 ‘방언’을 구사하거나 특정 인물의 성격을 구체화하는 것으로 다가왔다. 이외에 도 한국어의 음절언어적 특성을 드러내듯 음악의 흐름 안에 또렷하게 인지되는 고른 텍스트의 분 포가 존재했다. 그리고 이런 성부를 토대로 하는 대위법적인 짜임새의 표현성이 두드러졌다. 화성적으로 반음계 섹션을 등장시킴으로써 해당 구간에서 발화되는 텍스트를 독특한 방식으로 청취되도록 하거나, 실제 언어 수행 시 나타나는 상호작용적인 담화방식, 감정이 섞인 말의 반복, 특정 연령 집단이 보여주는 언어적 행동 등을 오페라 안에 음악적으로 구현한 경우도 있었다. 예컨대 이건용의 오페라는 한국어를 민감하게 다루고 있으며, 한국어의 다양한 음운론적인 특 징과 담화 상의 속성 등을 음악적으로 반영함으로써 표현력을 강화한다. 따라서 이건용의 오페라 는 개인의 오페라 작법을 구축한 것에서 한발 더 나아가, ‘한국오페라’로서의 기본적인 속성을 스 스로 구체화한 것으로도 볼 수 있을 것이다.

본 연구는 볼프강 아마데우스 모차르트의 초기 종교음악 중 ≪호칭기도≫ K. 109와 K. 125를 차용이라는 관점에서 고찰한다. 이 두 곡은 각각 1771년 5월, 1772년 3월에 잘츠부르크에서 작곡 되었는데, 아버지 레오폴트 모차르트의 영향을 받은 것으로 널리 알려져 있다. 하지만 구체적으로 그 영향이 어느 정도인지는 지금까지 논의되지 않았다. 본 연구에서는 모차르트의 곡들을 아버지 의 작품과 꼼꼼하게 비교하여 아버지의 영향이 가장 구체적인 경우에서부터 추상적인 경우까지 세 단계로 나누어 설명할 것이다. 모차르트가 아버지 레오폴트의 작품을 모델로 하여 어떻게 독 립적인 곡들을 만들어나갔는지 살펴보며, 이를 통해 모차르트의 ≪호칭기도≫ 뿐만 아니라 종교 음악 전반에 관한 학문적 관심을 불러일으키고자 한다.

Domestic nuclear power plants can affect the environment if multiple devices are operated on one site and even a trace amount of pollutants that may affect the environment after power generation are simultaneously discharged. Therefore, not only radioactive substances but also ionic substances such as boron should be discharged as minimally as possible. We adopted pilot CDI and SD-ELIX sytem to separating and concenrating of boron containing nulcear power plant discharge water. The boron concentration of the initial inflow water tended to decrease over time. The water quality of concentrated water also reached its peak until the initial 60 minutes, but tended to decrease in line with the decrease in the inflow water concentration. The boron removal rate was in the range of 85 to 99% with respect to the initial boron concentration of 15 to 25 mg/L. On the other hand, performance degradation due to the use of electrochemical modules is also observed, and regeneration through low ion-containing water cleaning effective. We shortened processing time by considering the optimal flow rate conditions and conductivity conditions and converting electrochemical modules into series or parallel.

Molten Salt Reactor, which employs molten salt mixture as fuel, has many advantages in reactor size and operation compared to conventional nuclear reactor. In developing Molten Salt Reactor, the behavior of fission product in operation should be preliminary evaluated for the correct design of reactor and its associated system including off-gas treatment. In this study, for 100 Mw 46 KCl- 54 UCl3 based Molten Salt Reactor with operating life time of 20 year, the fission product behavior was estimated by thermodynamic modeling employing FactSage 8.2. Total inventory of all fission product were firstly calculated using OpenMC code allowing depletion during neutronic calculation. Then, among all inventory, 46 element species from Uranium to Holmium were chosen and given to the input for equilibrium module of Factsage with its mass. In phase equilibrium calculation, for the correct description of solution phase, KCl-UCl3 solution database based on modified quasichemical model in the quadruplet approximation (ANL/CFCT-21/04) was employed and the coexisting solid phase was assumed to pure state. With the assumption of no oxygen and moisture ingress into reactor system, equilibrium calculation showed that 1% of solid phase and of gas phase were newly formed and, in gas phase, major species were identified : ZrCl4 (47%), Xe (33%), UCl4 (14%), Kr (5%), Ar (1%) and others. This result reveals that off-gas treatment of system should account for the appropriate treatment of ZrCl4 and UCl4 besides treatment of noble gas such as Xe and Kr.

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.

Exposure to lipopolysaccharide (LPS) causes cognitive impairment (CI). In the preliminary study, Lactobacillus gasseri NK109 suppressed LPS-induced expression of proinflammatory cytokines in macrophages. Therefore, the effect of NK109 on LPS-increased CI was investigated in mice. Intraperitoneal injection of LPS caused CI-like behaviors and neuroinflammation. However, orally administered NK109 reduced LPS-increased CI-like behaviors and hippocampal IL-1β and TNF-α expression, whereas LPS-decreased BDNF expression increased. NK109 also reduced LPS-increased colonic myeloperoxidase, IL-1β, and TNF-α expression. The efficacy of NK109 was increased by the combination of soybean embryo ethanol extract (SE). These findings suggest that NK109 with SE can improve CI by alleviating inflammation-mediated BDNF expression, thereby being beneficial for dementia therapy.

One of the most important factors in the delivery and acceptance requirements for dry storage of spent fuel is the burnup of spent fuel. Here, burnup has a unit of MWD/MTU and is used as a measure of how much nuclear fuel is depleted in a nuclear reactor. In addition, since it is one of the most basic characteristic information for the soundness evaluation of spent nuclear fuel, it is a required item not only by regulatory agencies but also by KORAD, the acquiring agency. The burnup of spent nuclear fuel is the burnup calculated through flux mapping using signals measured from in-reactor instruments during nuclear power plant operation (hereinafter: actual burnup) and the burnup calculated using the core design code (hereinafter: design burnup). In this paper, the design burnup of spent nuclear fuel discharged from OPR100 NPPs (Nuclear Power Plants) in Korea was recalculated to confirm the reliability of the actual burnup currently managed at the nuclear power plant. Basically, since spent nuclear fuel must maintain subcriticality under wet storage or dry storage, a burnup error of about 5% is considered as a conservative approach when evaluating the criticality safety of wet storage tanks and dry storage systems. Therefore, in this paper, we tried to verify whether the difference between actual burnup and design burnup for all spent nuclear fuel released from domestic OPR100 type light water reactor nuclear power plants is within 5%. As a result of the evaluation, the largest deviation between actual burnup and design burnup was about 1,457 MWD/MTU, and when converted into a percentage, it was about 3.3%. Therefore, it was confirmed that the actual burnup managed by OPR1000 NPPs in Korea has sufficient reliability. In the future, we plan to check the reliability of the performance burnup managed in WH NPPs, and some of them will be verified through measurement.