It is known that ZrCl4 can be used in the chlorination process of spent nuclear fuel. However, its solubility in high temperature molten salt is very small, making it difficult to dissolve a large amount of ZrCl4. Therefore, in this study, a flange-type sealed reactor was manufactured to observe the reaction characteristics of LiCl-KCl salt and ZrCl4. LiCl-KCl salt and ZrCl4 were placed in each alumina crucible, the reactor was sealed, and heated. The temperature at the reactor surface was above 500°C and maintained at that temperature for 48 hours. After completion of the reaction, the reactor was opened and the reaction products were recovered from each alumina crucible. The crystal structure of the reaction product was identified through XRD analysis, and the concentration of Zr was analyzed using ICP. Reaction characteristics were observed according to the molar ratio of ZrCl4 added to the number of moles of KCl in LiCl-KCl salt. The molar ratios of ZrCl4 to KCl were 0.5, 1, 2, and 3, respectively. As a result of each experiment, more than 95% of the injected ZrCl4 was vaporized and there was almost no residue in the ZrCl4 crucible. In the LiCl- KCl crucible, the weight increased in proportion to the amount of ZrCl4 added. As a result of XRD analysis, K2ZrCl6 was confirmed in all LiCl-KCl salt product. When the ZrCl4/KCl molar ratio was 2 and 3, LiCl-KCl could not be confirmed. Additionally, when the ZrCl4/KCl molar ratio was 1, LiCl was identified, but KCl was not found. Almost all of the KCl appears to have reacted with ZrCl4. ICP analysis results showed that the Zr concentration was proportional to the amount of ZrCl4 added in each LiCl-KCl salt, and exceeding the number of moles of reaction with KCl in the LiCl-KCl salt was observed. Therefore, these experimental results showed that ZrCl4 can be dissolved in LiCl-KCl salt at a maximum concentration higher than its solubility.

Th(IV) is a stable actinide that can act as a chemical analogue of U(IV) and Pu(IV), which are important radionuclides in safety assessments of deep geological repositories (DGR). Therefore, to understand the geochemical behaviour of U(IV) and Pu(IV), batch sorption of Th(IV) onto crystalline rocks were performed in oxidising conditions. The distribution coefficients (Kd) of Th(IV) were of particular interest. Gyeongju fresh groundwater (GF) and Gyeongju brackish groundwater (GB) were obtained at the Gyeongju Low and Intermediate Level Radioactive Waste (LILW) Disposal Facility. Crystalline granite (gr) and biotite gneiss (bg) were collected in Gyeongju and Gwacheon respectively and were grounded to a particle size smaller than 150 μm. Sorption samples were continuously shaken for 7 days under 200 rpm at 25°C. The liquid-to-solid ratio (V/m) was 200 L·kg-1. Th(IV) concentrations of the sorption samples were determined by UV-Vis-NIR absorption colorimetry from the formation of Th(IV)-arsenazo III complexes. Although the method allowed the initial Th(IV) concentrations to be determined, the final Th(IV) concentrations fell below the limit of detection (LOD), 6.27×10-9 mol·L-1. Taking the LOD as the final concentrations, conservative Kd were calculated to be 4,410 L·kg-1 for GF-gr and GF-bg, and 7,830 L·kg-1 for GB-gr and GB-bg. The result indicates a strong sorption affinity of Th(IV) onto granite and biotite gneiss within Gyeongju groundwater, suggesting a similar behaviour for U(IV) and Pu(IV). Furthermore, comparison of the conservative Kd obtained from the experiment were compared with existing Kd values of Th(IV). Such analysis and comparison of Th(IV) Kd in various types of groundwater could help locate the optimal site for a DGR in South Korea.

The safe disposal of high-level radioactive waste is a critical concern in many countries, especially in the context of the increasing use of nuclear power to overcome climate change. To provide a comprehensive understanding of the behavior of the radionuclides in the crystalline natural barrier, sorption of the artificially synthesized high-level radioactive waste (HLW) leachate was conducted. Granite (-1,000 m from ground level) and biotite gneiss (-100 m from ground level) rock cores were collected from Gyeongju and Gwacheon, respectively. The rock cores were milled with a jaw crusher and steel disk mill and then sieved. The crushed rocks with a diameter of 0.6 – 1.0 mm were selected, washed three times with deionized water, and then dried. To synthesize the simulated HLW leachate, representative elements (U(VI), Se(IV), Mo(VI), and Ni(II)) were added to natural groundwater collected from Gyeongju. The kinetic sorption experiment was performed in a polypropylene bottle with a solid-to-liquid ratio of 100 g/L in the orbital shaking incubator (200 rotations per min, 25.0°C). After the sorption, the supernatants were filtered by a 0.2-μm polytetrafluoroethylene syringe filter and subsequently analyzed by inductively coupled plasma-mass spectrometry (ICP-MS). Through the kinetic change of aqueous concentration, the contact time has been determined to be 7 days. Ni(II) showed the highest distribution coefficients (Kd = 0.81 L/m2 for granite and 8 – 16 L/m2 for biotite gneiss), followed by U(VI) (Kd = 0.03 – 0.04 L/m2 for granite and 0.04 – 0.05 L/m2 for biotite gneiss). Highly mobile nuclides such as Se(IV) (Kd = 0.02 L/m2 for granite and 0.03 L/m2 for biotite gneiss) and Mo(VI) (Kd = 0.01 – 0.02 L/m2 for granite and 0.01 L/m2 for biotite gneiss) showed the lowest distribution coefficient. Our study provides insights into the migration-retention behaviors of the HLW leachate with granite and biotite gneiss in geological systems and verifies the sorption parameters, e.g., distribution coefficients, experimentally produced by other groups to ensure the safe disposal of HLW.

As a method for chlorinating spent nuclear fuel, a method of using ZrCl4 in high-temperature molten salt is known. However, ZrCl4 has a sublimation property that vaporizes at a temperature similar to the melting temperature of molten salt. Since solubility of ZrCl4 in molten salt is very low, it is difficult to dissolve a large amount of ZrCl4 in molten salt. However, once ZrCl4 can be dissolved together with the molten salt, it remains in the molten salt without vaporizing. That is, it is known that when vaporized ZrCl4 reacts with molten salt in a sealed reactor, it dissolves into the molten salt, and ZrCl4 above the solubility remains in the molten salt in the form of M2ZrCl6. Here, M represents an alkali element. Therefore, in this study, a flange-type sealed reactor was fabricated to dissolve a large amount of ZrCl4 in LiCl-KCl salt, and LiCl-KCl salt in which ZrCl4 was dissolved as K2ZrCl6 was prepared. LiCl-KCl, KCl, and ZrCl4 salts were charged into alumina crucibles and placed in a sealed reactor. The reactor was heated to 500°C and the reaction time was about 20 hours. The temperature of the reactor surface was about 480°C. After completion of the reactions, each crucible was recovered from the inside of the reactor. All of the ZrCl4 vaporized and there was no residue in the crucible. Both KCl and LiCl-KCl salts appear to have dissolved and then cooled, with respective weight gains. XRD analysis was performed to observe the structure of the recovered salts, and ICP analysis was performed to measure the Zr element content in each salt. As a result of XRD analysis, the structure of K2ZrCl6 was found in the KCl salt, but not in the LiCl-KCl salt. As a results of ICP analysis, it was found that the LiCl-KCl salt contained about 33wt% of ZrCl4, and about 25wt% was dissolved in the KCl salt. In other words, it was shown that ZrCl4 above the solubility can be dissolved in the LiCl-KCl molten salt.

To estimate the removal efficiency of TRU and rare earth elements in an oxide spent fuel, basic dissolution experiments were performed for the reaction of rare earth elements from the prepared simfuel with chlorination reagents in LiCl-KCl molten salt. Based on the literature survey, NH4Cl, UCl3, and ZrCl4 were selected as chlorination reagent. CeO2 and Gd2O3 powders were mixed with uranium oxide as a representative material of rare earth elements. Simfuel pellets were prepared through molding and sintering processes, and mechanically pulverized to a powder form. The experiments for the reaction of the simfuel powder and chlorination reagents were carried out in a LiCl-KCl molten salt at 500°C. To observe the dissolution behavior of rare earth elements, molten salt samples were collected before and after the reactions, and concentration analysis was performed using ICP. After the reaction completed, the remaining oxide was washed with water and separated from the molten salt, and XRD was used for structural analysis. As a result of salt concentration analysis, the dissolution performance of rare earth elements was confirmed in the reaction experiments of all chlorination reagents. In an experiment using NH4Cl and ZrCl4, the uranium concentration in the molten salt was also measured. In other words, it seemed that not only rare elements but also uranium oxide, which is a main component of simfuel, was dissolved. Therefore, it is thought that the dissolution of rare earth elements is also possible due to the collapse of the uranium oxide structure of the solid powder and the reaction with the oxide of rare earth elements exposed to molten salt. As a result of analyzing the concentration changes of Simfuel before and after each reaction, there was little loss of uranium and rare earth elements (Ce/Gd) in the NH4Cl experiment, but a significant amount of rare earth elements were found to be reduced in the UCl3 experiment, and a large amount of rare earth elements were reduced in the ZrCl4 reaction.

The concentrations of four potentially toxic elements were analyzed in 70 spraint samples of Eurasian otters from the Daechung Lake area. The median concentrations of As, Hg, Ni, and Pb were 1.19, 1.06, 9.95, and 0.37 μg/g dry weight, respectively. A moderate negative correlation was observed between the levels of Hg and Pb (rs = -0.47). The Hg levels showed a weak negative correlation with the distances of the sampling locations from the Daechung Dam (rs = -0.35). This study provides baseline information on the health risks associated with the hazardous elements in otter species.

희토류(Rare Earth) 함량이 높은 TRU 생성물 중의 RE 원소를 감소시키기 위하여 RE 원소와 UCl3의 산화반응을 이용한 RE 제거 공정의 타당성을 HSC Chemistry 코드를 이용하여 검토하였다. 사용후핵연료에 포함된 TRU 원소 및 RE 원소의 조성 및 열역학적 자료를 검토하였으며, UCl3와의 산화 반응에 따른 평형 자료를 계산하여 공정 가능성을 검토하였다. 실제 파이로 프로세싱 처리를 가정한 물질수지로부터 TRU 생성물의 RE 함량이 다른 두 가지 경우에 대하여 RE 원소 제거율과 TRU 회수율을 평가하였다. TRU 생성물을 산화제인 UCl3와 반응시켰을 때 각 원소의 Gibbs free energy의 차이에 의한 선택적 산화 반응이 일어났다. 투입된 UCl3 양을 조절하여 TRU 회수율을 최대로 유지하면서도 RE 원소를 제거하여 최종생성물의 TRU/ RE 비를 증가시킬 수 있는 가능성을 계산 결과로 확인하였다. 본 연구의 결과들은 열역학적 평형 자료에 기반한 결과이므로 실제 공정에 적용하기에는 많은 차이점이 존재한다. 그러나 TRU 물질을 취급하기 어려운 환경에서 파이로프로세싱의 TRU 생성물을 고속로의 핵연료로 공급하기 위한 공정 설계에 중요한 자료로 활용될 것으로 기대된다.

본 연구에서는 다물체 페리다이나믹 해석 코드의 MPI-OpenMP 혼합 병렬화를 수행하였다. 페리다이나믹 해석 모델은 복잡한 동적 파괴 거동 및 불연속 특성을 모사하는데 적합하지만, 비국부 영역을 통한 절점 간 상호작용을 계산하기 때문에 유한요소 모델에 비해 계산 시간이 많이 소요된다. 또한 다중적층구조물의 다물체 페리다이나믹 해석에서 추가된 비국부 접촉 모델과 가상 층간 결합 모델을 통한 여러 물체 간 상호작용으로 계산 부담이 증가한다. 더불어 고속 충돌 파괴와 같은 복잡한 동적 파괴 거동 해석을 위해 세밀한 절점 간격과 작은 시간 간격이 요구되기 때문에 코드 최적화와 병렬화를 통한 고성능 해석 코드 개발이 필수적이다. 해석 코드는 Intel Fortran MPI compiler와 OpenMP를 사용하여 개발되었으며, 한국과학기술정보원(KISTI)의 슈퍼컴퓨팅센터 누리온(Nurion)으로 실행되었다. 다물체 해석 코드를 최적화하기 위한 핵심 요소들을 분석하고, 모델 의존성 발생 서브루틴 분석 및 프로세스 통신 데이터 분별을 통해 MPI-OpenMP 혼합 병렬 처리 구조를 적용하였다. 다물체 충돌 파괴 현상 시뮬레이션을 통해 개발된 병렬 처리 코드의 성능을 확인하였다.

파이로프로세싱의 전해환원공정에서 생산된 금속전환체의 조성은 전해정련공정 운전의 중요한 운전변수인 용융염 중 UCl3 의 농도변화에 영향을 미친다. 따라서, 본 연구에서는 금속전환체에 함유된 TRU와 RE 원소의 함량 및 금속전환체에 동반 되어 전해정련 전해조에 유입될 수 있는 Li2O 농도가 전해정련 전해조의 UCl3 농도 변화에 미치는 영향을 검토하였다. 금속 전환체의 TRU 원소와 RE 원소의 농도만을 고려하였을 때 전해정련 운전 batch 수 증가에 따라 UCl3 농도가 감소하였다. 전 해정련 1 campaign(20 batch)를 운전하기 위해서는 UCl3를 3회 이상 추가 보충해야 함을 알 수 있었다. 한편, 금속전환체에 동반되어 전해정련 전해조에 유입되는 Li2O의 유입량 증가에 따라 UCl3 농도 감소의 영향이 크게 나타났으며, 이에 따라 운전 가능 batch 수가 급격히 감소하게 되어 전해정련 운전에 중요한 운전 변수임을 보여주었다. 이러한 결과는 전해정련 운전 중 UCl3 농도 유지를 위해 금속전환체에 포함된 TRU 및 RE 원소뿐만 아니라 금속전환체에 동반되어 유입될 가능성이 있는 Li2O의 영향도 고려하여 전해정련 운전모드를 설정하여야 함을 보여주었다.

In the real world, most of biological systems that follow Markov process have internal states which are unobservable, so called hidden states. However, although the states could not be directly observed, events emitted under any hidden states would often be observable. It infers that a lot of biological systems might be simulated by hidden Markov model (HMM). To date, many studies tried to apply the HMM for monitoring and describing of animal behaviors. In this study, we attempted to build a HMM that emulates the traveling behavior of subterranean termites in the artificial tunnels with varying width and curvature, based on the empirical data obtained from our previous study that explored the relationship between subterranean termite's movement efficiency and the geometric pattern of their tunnels.