음이온 교환막(AEM) 수전해용 AEM 소재 개발은 재생 에너지를 활용한 수소 생산 기술을 개선하는 데 중요한 역할을 한다. 이러한 소재를 설계하고 최적화하는 데 분자동역학 전산모사가 유용하게 사용되지만, 전산모사 결과의 정확도 는 사용된 force-field에 크게 의존한다. 본 연구의 목적은 AEM 소재의 구조와 이온 전도 특성을 예측할 때 force-field 선택 이 미치는 영향을 체계적으로 조사하는 것이다. 이를 위해 poly(spirobisindane-co-aryl terphenyl piperidinium) (PSTP) 구조를 모델 시스템으로 선택하고 COMPASS III, pcff, Universal, Dreiding 등 네 가지 주요 force-field를 비교 분석하였다. 각 force-field의 특성과 한계를 평가하기 위해 298~353 K의 온도 범위에서 수화 채널 형태, 물 분자와 수산화 이온의 분포, 수산 화 이온 전도성을 계산하였다. 이를 통해 AEM 소재의 분자동역학 전산모사에 가장 적합한 force-field를 제시하고, 고성능 AEM 소재 개발을 위한 계산 지침을 제공하고자 한다.

본 연구에서는 실리카 복합막 기반 고분자 전해질막을 5단 연료전지 스택에 적용하여 성능 평가를 수행하였다. 이를 통하여, 개별 구성 요소의 성능도 중요하지만, 전체적인 관점에서 공급되는 연료의 유량이 스택 성능에 중요한 역할을 하며, 특히 수소의 유량에 크게 의존한다는 사실이 확인하였다. 산소의 유량을 증가시켜도 성능의 변화는 미미한 반면, 수소 의 유량을 증가시키면 성능이 향상되는 것을 확인하였다. 그러나 수소의 유량 증가는 수소와 산소 유량 비율의 불균형을 초 래하여 장기적으로는 스택 성능과 내구성을 저하시키는 문제가 관찰되었다. 이러한 현상을 스택 구성 요소 및 개별 단위 셀 에서도 관찰할 수 있었으며, 따라서 스택 운전 시 각 구성 요소의 성능을 최적화하는 것 외에도 균일한 유량 제어를 위해 유 로 설계 및 운전 조건을 최적화하는 것이 중요하다는 것을 알 수 있었다. 마지막으로 실리카 복합막은 최대 출력 기준 25 W 이상의 성능을 나타내어 실제 연료전지 시스템에 적용하기에 충분한 성능을 갖춘 것으로 판단된다.

국내에서 분리된 Bacillus thuringiensis NT0423은 나비목과 파리목 곤충에 독성을 보이는 130 kDa의 전형적인 다이아몬드형 내독소 단백질을 생성한다. 이 균주의 파리목에 대한 독성을 강화하기 위하여, B. thuringeinsis NT0423에 모기 유충에 강한 독성을 보이는 B. thuringiensis subsp. morrisoni PG-14의 cryVID유전자를 가지고 있는 pCG10 플라스미드를 electroporation 방법을 이용하여 형질전환하였다. 형질전환체인 B. thuringiensis PT1227내에서 cryIVD와 숙주가 생성하는 원래의 다이아몬드형 130kDa 내독소 단백질 유전자는 그 자신의 형태로 잘 발현되었다. 형질전환체의 모기 유충에 대한 독성은 원래 숙주의 내 독소 단백질과ㅏ 도입된 CryIVD의 상승효과에 의해 현저히 증가하였다.

Radioactive iodine-129, a byproduct of nuclear fission in nuclear power plants, presents significant environmental and health risks due to its high solubility in water and volatility. Iodine-129, with its half-life of 1.57×1017 years, necessitates safe management and disposal. Therefore, safely capturing and managing I-129 during spent nuclear fuel reprocessing is of paramount importance. To address these challenges, various glass waste forms containing silver iodide have been developed, such as borosilicate, silver phosphate, silver vanadate, and silver tellurite glasses. These glasses effectively immobilize iodine, but the high cost of silver raises affordability concerns. This study introduces CuI·Cu2O·TeO2 glass waste forms for iodine immobilization, a novel approach. The cost-effectiveness of copper, in contrast to silver, makes it an attractive alternative. The CuI·Cu2O·TeO2 glass waste forms were synthesized with varying CuI content (x) in (1-x)(0.3Cu2O·0.7TeO2) glass matrices. Xray diffraction (XRD) confirmed amorphous structures, and X-ray fluorescence (XRF) quantified composition. X-ray photoelectron spectroscopy (XPS) and Raman spectroscopy provided insights into structural properties. Durability assessments using a 7-day product consistency test (PCT-A) and inductively coupled plasma-mass spectrometry (ICP-MS) revealed compliance with U.S. glass regulations, making CuI·Cu2O·TeO2 glasses a promising choice for iodine immobilization in radioactive waste.

The nuclide management process for reducing the environmental burden being developed by the Korea Atomic Energy Research Institute is performed in molten salts, resulting in contaminated salt wastes containing fission products such as Cs, Sr, Ba, and rare-earth nuclides. In addition, the spent fuel of a molten salt reactor (MSR) contains a variety of fission products, and a purification process may be required for the reuse of the salt and the separation and disposal of the fission products in the spent nuclear fuel. The melt-crystallization method is a technique used for the purification and separation of chemicals or metals based on the different melting points of the different substances. In a recent study, our group developed a reactive-crystallization method using Li2CO3 precipitation agent to precipitate metal corrosion from the reactor through a chlorination reaction by HCl and Cl2, which may occur in chloride molten salt, and successfully precipitated the metal precipitate and purified and recovered LiCl salt. In this study, reactive-crystallization method has been established for removing fission products and corrosive materials. Using the reactive crystallization method, white LiCl-KCl salt that was not discolored by metal corrosion was recovered through the crystallization plates, and fission products and metal elements were shown to be suppressed to several ppm in the purified salt. Consequently, high-purity salts were recovered with high nuclide and corrosive separation efficiencies. The reactive crystallization procedure can also be applied to other salt waste systems, such as MSR nuclear fuel treatment and molten salt chemistry for the elimination of corrosive substances.

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.

Heat-generating nuclides such as Cs-137 and Sr-90 should be separated from spent nuclear fuel to reduce the short-term thermal load on the repository facility. In particular, Sr-90 must be separated because its decay process generates high temperatures. Recently, the Korea Atomic Energy Research Institute (KEARI) has been developing a waste burden minimization technology to reduce the environmental burden resulting from the disposal of spent nuclear fuel and maximize the utilization of the disposal facility. The technology incorporates a nuclide management process that maximizes disposal efficiency by selectively separating and accumulating key nuclides from spent nuclear fuel, such as Cs, Sr, I, TRU/RE, and Tc/Se. Sr nuclides dissolve in the chloride phase during the chlorination process of spent nuclear fuel and are recovered as carbonate or oxide through reactive distillation or reactive crystallization. Due to their chemical similarity, Ba nuclides are recovered along with Sr nuclides during this process. In this study, we prepared a ceramic waste form for group II nuclides, Ba(x)Sr(1-x)TiO3 (x=0, 0.25, 0.5, 0.75, 1), using the solid-state reaction method, taking into account the different ratios of Sr/Ba nuclides produced during the nuclide management process. Regardless of the Sr/Ba ratio, the established waste form fabrication process was able to produce a stable waste form. Physicochemical properties, including leaching and thermal properties, were evaluated to determine the stability of group II waste forms. In addition, the radiological properties of waste forms of Ba(x)Sr(1-x)TiO3 with varying Sr/Ba ratios were evaluated. These results provided fundamental data for the long-term storage and management of waste forms containing group II nuclides.

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.

Se-79, a fission product of uranium, is present in spent nuclear fuel. Selenium is volatilized from the spent nuclear fuel during the pretreatment of pyroprocessing, and a filter composed of calcium oxide can capture gaseous selenium in the form of CaSeO3. Because Se-79 has a long half-life (3.27E5 years) and selenite ions have high mobility in groundwater, they must be immobilized in a chemically stable form for final disposal. This study used a composition of 50 TeO2 - 10 Al2O3 - 10 B2O3 - 10 Na2O - 10 CaO - 10 ZnO (mol%). High-purity powders of TeO2, Al2O3, H3BO3, Na2CO3, CaCO3, and ZnO were used as glass precursors. The mixed powders were placed in alumina crucibles and melted in an electric furnace under an ambient atmosphere at 800°C for 1 h before being cast on a carbon mold. The obtained glasses were ground into fine powders and then mixed with CaSeO3 powders. The powders were melted in alumina crucibles at 800°C for 1 h. To simulate a seleniumcaptured calcium filter, CaSeO3 was synthesized by a precipitation method using sodium selenite (Na2SeO3) and calcium nitrate (Ca(NO3)2) solutions. The glass samples were analyzed by an X-ray diffractometer (XRD). Retention of Se in tellurite glasses was analyzed by an X-ray fluorescence spectrometer (XRF) and inductively coupled plasma (ICP). The chemical durability of tellurite glass was evaluated through the PCT method.