산오이풀(Sanguisorba hakusanensis)은 한국의 자생식물 이며 정원소재로써 가치가 있지만, 생육 및 생리적 특성 및 정 원 적응 여부에 대하여 알려진 정보가 많지 않아 이용에 어려 움을 겪고 있다. 본 연구에서는 자생식물인 산오이풀의 관수 주기 및 NaCl 농도에 따른 생장, Fv/Fm, NPQ, 성분 변화, 무기성분 변화를 조사하여 내건 및 내염성 보유 여부, 생육 한 계 범위, 스트레스 환경에서 생육을 유지하기 위한 반응을 파 악하고자 했다. 실험 결과 NaCl 무처리구의 관수주기에 따른 성분 분석에서 엽록소 함량의 감소를 제외하고 유의한 차이가 나타나지 않았으나 이는 토양수분함량이 건조 스트레스를 유 발할 정도로 감소하지 않았기 때문으로 보인다. 염 처리에서 는 2주 이후 급격한 스트레스 반응이 나타났고 3주차부터 고 사하기 시작하여 6주차에 모든 개체가 최종 고사했다. 이러한 결과는 2주까지 염 스트레스에 의해 유발되는 2가지 스트레 스 중 초기에 나타나는 삼투 스트레스에는 저항하였으나 이후 나타나는 NPQ의 감소 등 이온 스트레스에 의해 유발된 광합 성 기구 붕괴로 인해 정상적인 생육을 유지할 수 없었기 때문 에 나타난 것으로 보인다. 그러나 무기이온 분석은 이온 스트 레스에 저항하기 위한 메커니즘의 존재 가능성을 시사하였다. 상대적으로 염 농도가 낮을 때에는 세포내 Ca2+ 및 K+ 수준이 높았는데, 이는 Ca2+ 수준이 높아짐에 따라 Na+를 세포 밖으 로 방출시키는 단백질, Na+를 K+와 함께 수송하는 단백질이 기능하여 Na+축적을 지연시키는 반응이 있었음을 시사한다. 그러나 NaCl을 고농도로 처리했을 때는 이러한 반응이 관찰 되지 않았다. 따라서 산오이풀은 염 스트레스에 의해 야기되 는 삼투 스트레스에 강한 저항성을 가지고 있고 이온 독성을 줄이기 위한 메커니즘으로 Na+ 세포내 축적을 지연시키는 것으로 보이지만, 심한 염 스트레스를 받았을 때 나타나는 급격 한 반응에서 이러한 메커니즘이 기능하지 못하고 이온독성에 매우 취약한 것으로 여겨진다. 본 연구를 통해 자생식물인 산 오이풀의 활용을 늘리는 데 기초적인 자료를 제공할 수 있을 것으로 생각된다.

Wolsong Unit 1, a domestic heavy water reactor nuclear power plant, was permanently shut down in December 2019. Accordingly, Wolsong Unit 1 plans to prepare a Final Decommissioning Plan (FDP), submit it to the government by 2024, receive approval for decommissioning, and begin full-scale decommissioning. One of the important tasks in the decommissioning of Wolsong Unit 1 is to determine the decommissioning strategy. It is necessary to decide on a decommissioning strategy considering various factors and variables, secure the technical background, and justify it. The selection of a decommissioning strategy is best achieved through the use of formal decisionmaking assistance techniques, such as considerations related to influencing factors. It is very important to understand the basic decommissioning strategy alternatives and whether sufficient consideration has been given to situations where only a single unit is permanently shut down in a multi-unit site like Wolsong Unit 1, while the remaining units are in normal operation. As a process for selecting a decommissioning strategy, first, all considerations that could potentially affect decommissioning presented in the KINS Decommissioning Safety Review Guidelines were synthesized, influencing factors to be used in the decision-making process were determined, and the concept was defined. In order to select the most appropriate decommissioning strategy by considering various evaluation attributes of possible decommissioning alternatives (immediate dismantling and delayed dismantling), the Wolsong Unit 1 decommissioning strategy was evaluated by reflecting the AHP decision-making technique.

표면발현(surface-display system)은 세포 또는 바이러스 표면에 목적 단백질을 고정하여 발현시킴으로써 목적 단백질에 대하여 독립적인 공간 구조 및 생물학적 활성을 부여하는 단백질 공학 기술이다. 또한 이를 이용하여 높은 중화항체 유도 및 대량생산이 가능한 삼량체의 형태로 항원 단백질의 발현 또한 가능하다. BES(baculovirus expression system)에서의 표면발현 기술은 번역 후 수정과정 및 복잡한 구조의 다양한 단백질의 발현이 가능하기 에 다른 숙주 기반 시스템보다 효율적이라고 보고되고 있다. 그러나 목적 단백질 외의 다른 표면 단백질과 발현 공간에서의 경쟁으로 목적 단백질의 낮은 생산량이 큰 문제점으로 지적되고 있다. 따라서, 이러한 BES에서 표면 발현의 생산 효율을 증대시키기 위하여, 동일한 표면 공간에 대한 단백질 간의 발현 경쟁에 대해 실험적으로 확인 후, 그를 해결하기 위하여 표면발현에 최적인 목적 단백질 발현을 위한 프로모터 선발 실험을 수행하였다. 이를 통해 BES에서 표면발현에 의한 목적 단백질의 생산 효율을 증대시킬 수 있음을 확인하였다.

Boric acid-containing B-10 is used in a nuclear reactor as a coolant and absorbs thermal neutrons generated during nuclear fission in the primary circuit. Boron-containing coolant water waste is generated from maintenance, floor drain, decontamination, and reactor letdown flows. There are two options for aqueous solution waste of boric acid. One is recycling and discharge through filtration, ion exchange, and reverse osmosis. The other is immobilization after evaporation and crystallization processes. The dry powder of boric acid waste liquid can be immobilized by cement, polymer, etc. Before the mid-1990s, concentrated boric acid waste was solidified with a cement matrix. To overcome the disadvantage of low waste loading of cement waste form, a method of solidifying with paraffin was adopted. However, paraffin solids were insufficient to be disposed of as final waste. Paraffin is a kind of soft solidified material and has low compressive strength and poor leaching resistance. As a result, it was decided as an unsuitable form for disposal. In KOREA, paraffin waste form was adopted for boric acid waste treatment in the 1990s. A large amount of paraffin waste forms about 20,000 drums (200 l drum) were generated to treat boric acid waste and were stored in nuclear power sites without disposal. In this study, we want to obtain high-purity boric acid waste by oxidizing and decomposing solid paraffin waste form through a boric acid catalytic reaction. In this reaction, paraffin is separated in the form of various by-products, which can then be treated through a liquid waste treatment device or an exhaust gas treatment device. The proper temperature for sample decomposition during the catalytic reaction was set through TGA analysis. Compositions of by-products and residues generated at each stage of the reaction could be analyzed to determine the state during the reaction. Finally, the boric acid waste powder was perfectly separated from paraffin waste form with disposable products through this pyrolysis process.

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.