본 연구는 신생아 검사 중 포수클로랄(chloral hydrate)을 투여 후 진행되는 신생아 진정 검사 대비 진정 대체 방식 중 하나인 피드 및 랩(feed and wrap) 방식의 유용성을 평가한 연구이다. 본 연구에선 진정으로 진행한 신생아의 두뇌 T2 축면 영상과 피드 및 랩 방식으로 진행한 같은 영상 각 30개의 운동 허상(motion artifact)과 백질과 회백질의 구분 정도를 두 명의 영상의학과 전문의가 정성적으로 평가하였고, 운동 허상을 측정하기 위해서 위상부호화(phase encoding) 방향의 배경 영역(background area)의 평균 신호 강도(mean signal intensity)를 구하여서 정량적 방식으로 평가하였다. 또한 총검사 시간을 정리한 뒤 정량적 방식으로 평가하였고 투약 기록의 여부와 간호일지를 토대로 피드 및 랩 방식의 총 39건의 검사 건수 대비 성공률을 측정하였다. 운동 허상의 정량적 평가와 영상 품질의 정성적 평가 모두에서 두 집단은 유의미한 차이가 없었으나, 검사 시간의 정량적 평가에선 p값이 0.001로 유의한 차이가 있었다. 피드 및 랩 방식의 총검사 건수 대비 성공률은 100%였다. 결론적으로 본 논문에선 피드 및 랩 방식과 진정 방식의 영상 품질이 유의한 차이가 없고 성공률이 높기에 유용하다고 판단하였으나, 검사 시간이 더 지연되는 한계가 있다는 사실을 확인하였다.

A low- and intermediate-level radioactive waste repository contains different types of radionuclides and organic complexing agents. Their chemical interaction in the repository can result in the formation of radionuclide-ligand complexes, leading to their high transport behaviors in the engineered and natural rock barriers. Furthermore, the release of radionuclides from the repository can pose a significant risk to both human health and the environment. This study explores the impact of different experimental conditions on the transport behaviors of 99Tc, 137Cs, and 238U through three types of barrier samples: concrete, sedimentary rock, and granite. To assess the transport behavior of the samples, the geochemical characteristics were determined using X-ray diffraction (XRD), X-ray fluorescence (XRF), Fouriertransform infrared spectroscopy (FTIR), scanning electron microscopy with energy-dispersive X-ray spectroscopy (SEM-EDS), and Brunauer-Emmett-Teller (BET) analysis. The adsorption distribution coefficient (Kd) was used as an indicator of transport behavior, and it was determined in batch systems under different conditions such as solution pH (ranging from 7 to 13), temperature (ranging from 10 to 40°C), and with the presence of organic complexing agents such as ethylenediaminetetraacetic acid (EDTA), nitrilotriacetic acid (NTA), and isosaccharinic acid (ISA). A support vector machine (SVM) was used to develop a prediction model for the Kd values. It was found that, regardless of the experimental parameters, 99Tc may migrate easily due to its anionic property. Conversely, 137Cs showed low transport behaviors under all tested conditions. The transport behaviors of 238U were impacted by the order of EDTA > NTA> ISA, particularly with the concrete sample. The SVM models can also be used to predict the Kd values of the radionuclides in the event of an accidental release from the repository.

The organic complexing agents such as ethylenediaminetetraacetic acid (EDTA), nitrilotriacetic acid (NTA), and isosaccharinic acid (ISA) can enhance the radionuclides’ solubility and have the potential to induce the acceleration of radionuclides’ mobility to a far-field from the radioactive waste repository. Hence, it is essential to evaluate the effect of organic complexing agents on radionuclide solubility through experimental analysis under similar conditions to those at the radioactive waste disposal site. In this study, five radionuclides (cesium, cobalt, strontium, iodine, and uranium) and three organic complexing agents (EDTA, NTA, and ISA) were selected as model substances. To simulate environmental conditions, the groundwater was collected near the repository and applied for solubility experiments. The solubility experiments were carried out under various ranges of pHs (7, 9, 11, and 13), temperatures (10°C, 20°C, and 40°C), and concentrations of organic complexing agents (0, 10-5, 10-4, 10-3, and 10-2 M). Experimental results showed that the presence of organic complexing agents significantly increased the solubility of the radionuclides. Cobalt and strontium had high solubility enhancement factors, even at low concentrations of organic complexing agents. We also developed a support vector machine (SVM) model using some of the experimental data and validated it using the rest of the solubility data. The root mean square error (RMSE) in the training and validation sets was 0.012 and 0.016, respectively. The SVM model allowed us to estimate the solubility value under untested conditions (e.g., pH 12, temperature 30°C, ISA 5×10-4 M). Therefore, our experimental solubility data and the SVM model can be used to predict radionuclide solubility and solubility enhancement by organic complexing agents under various conditions.

The dry storage of spent fuel has become an increasingly important issue in the field of nuclear energy. Square-gridded baskets have been widely used for the storage of spent fuel because of their superior heat transfer and structural integrity. In this paper, we review the fabrication process of square-gridded baskets for dry storage of spent fuel. The review includes the design considerations, material selection, manufacturing methods, and quality control measures. We also discuss the challenges and opportunities for further improvement in the fabrication of square-gridded baskets. The fabrication of square-gridded baskets is a critical process for the safe and reliable dry storage of spent fuel. The review of the fabrication process highlights the importance of design considerations, material selection, manufacturing methods, and quality control measures. Continued efforts to improve the fabrication process will help to ensure the safe and secure storage of spent fuel.

As temporary storage facilities for spent nuclear fuel (SNF) are becoming saturated, there is a growing interest in finding solutions for treating SNF, which is recognized as an urgent task. Although direct disposal is a common method for handling SNF, it results in the entire fuel assembly being classified as high-level waste, which increases the burden of disposal. Therefore, it is necessary to develop SNF treatment technologies that can minimize the disposal burden while improving long-term storage safety, and this requires continuous efforts from a national policy perspective. In this context, this study focused on reducing the volume of high-level waste from light water reactor fuel by separating uranium, which represents the majority of SNF. We confirmed the chlorination characteristics of uranium (U), rare earth (RE), and strontium (Sr) oxides with ammonium chloride (NH4Cl) in previous study. Therefore, we prepared U-RE-SrOx simulated fuel by pelletizing each elements which was sintered at high temperature. The sintered fuel was again powdered by heating under air environment. The powdered fuel was reacted with NH4Cl to selectively chlorinate the RE and Sr elements for the separation. We will share and discuss the detailed results of our study.