리튬이온전지는 친환경적이고 우수한 전지 성능덕분에 배터리 산업의 핵심으로 자리 잡았으며, 이에 따라 수요가 급증하고 있다. 그러나, 리튬이온전지의 수요증가는 리튬과 광물자원들의 공급문제를 초래하며, 수명이 다한 폐 리튬이온전지의 폐기방안이 아직 마련되지 않아 환경적 문제를 발생시킨다. 이러한 문제를 해결하기 위해 폐 리튬이온전지를 재활용하는 연구가 진행되고 있으며, 그 중에서도 폐 리튬이온전지에서 폐 양극 소재를 추출하여 재활용하는 다이렉트 리사이클링 연구가 주목받고 있다. 그러나, 폐 양극 소재는 오랜 충/방전으로 인해 구조적 붕괴(열화)가 발생한 상태로, 새로운 리튬이온전지에 적용을 위해서는 리튬이온전지 사용 전의 구조 즉, 층상구조로의 회복이 필요하다. 본 연구에서는 이를 위해 폐 양극 소재(LiNi0.6C0.2Mn0.2O2)가 열역학적으로 층상구조를 형성하는 온도를 분석하기 위해 700 ºC, 800 ºC, 900 ºC 범위에서 XRD를 통해 구조분석을 진행하였다. 폐 양극 소재는 700 ºC와 900 ºC 대비 800 ºC 열처리 시 1.44로 가장 높은 I003/I104 value를 보였다. 또한 800 ºC 열처리 시 0.1 C 기준 비 용량이 171.3 mAh/g으로 가장 높은 것을 확인하였다. 이를 통해 우리는 열역학적으로 층상구조를 형성하는 온도를 800 ºC로 도출하였으며 폐 양극 소재의 구조를 성공적으로 복원하였다.

In Korea, two types of spent nuclear fuels (SNFs) are generated, pressurized light water reactor type (PWR) and pressurized heavy water reactor type (PHWR; CANDU), that differ greatly in size, decay heat, and radioactive characteristics. Technology development for the disposal of SNFs has mainly focused on PWR SNFs that are large in size and have extremely high decay heat and radioactivity. However, CANDU SNFs should be considered differently from PWR SNFs in deep geological disposal systems because their characteristics significantly differ from those of PWR SNFs in terms of their dimensions, number of SNF bundles, and handling systems in nuclear power plant sites. In this paper, after reviewing the status of the CANDU SNF disposal concept by Canada and Korea, concepts related to the direct geological disposal of CANDU SNFs were described, and two concepts were proposed based on the results of the development. The engineered barrier systems developed using these two concepts were comparatively analyzed in terms of disposal safety, disposal efficiency, and technical maturity. Based on the results of the comparative analyses, a vertical-type emplacement disposal concept was determined as a reference concept for the deep geological disposal of CANDU SNFs.

The spent nuclear fuel, combusted and released in the nuclear power plant, is stored in the spent fuel pool (SFP) located in the fuel buildings interconnected with the reactors. In Korea, spent fuel has been stored exclusively in SFPs, prompting initiatives to expand storage capacity by either installing additional SFPs or replacing them with high-density spent fuel storage racks. The installation of these fuel racks necessitates obtaining a regulatory license contingent upon ensuring safe fuel handling and storage systems. Regulatory agencies mandate the formulation of various postulated accident scenarios and assessments covering criticality, shielding, thermal behavior, and structural integrity to ensure safe fuel handling and storage systems. This study describes an evaluation method for assessing the structural damage to storage racks resulting from fuel dropping as a part of the functional safety evaluation of these racks. A scenario was envisaged wherein fuel was dropped onto the base plates of the upper and lower sections of the storage racks, and the impact load was analyzed using the ABAQUS/Explicit program. The evaluation results revealed localized plastic deformation but affirmed the structural integrity and safety of the storage racks.

Hydride analysis is required to assess the mechanical integrity of spent nuclear fuel cladding. Image segmentation, which is a hydride analysis method, is a technique that can analyze the orientation and distribution of hydrides in cladding images of spent nuclear fuels. However, the segmentation results varied according to the image preprocessing. Inaccurate segmentation results can make hydride difficult to analyze. This study aims to analyze the segmentation performance of the Otsu algorithm according to the morphological operations of cladding images. Morphological operations were applied to four different cladding images, and segmentation performance was quantitatively compared using a histogram, betweenclass variance, and radial hydride fraction. As a result, this study found that morphological operations can induce errors in cladding images and that appropriate combinations of morphological operations can maximize segmentation performance. This study emphasizes the importance of image preprocessing methods, suggesting that they can enhance the accuracy of hydride analysis. These findings are expected to contribute to the advancements in integrity assessment of spent nuclear fuel cladding.

Because most spent nuclear fuel storage casks have been designed for low burnup fuel, a safety-significant high burnup dry storage cask must be developed for nuclear facilities in Korea to store the increasing high burnup and damaged fuels. More than 20% of fuels generated by PWRs comprise high burnup fuels. This study conducted a structural safety evaluation of the preliminary designs for a high burnup storage cask with 21 spent nuclear fuels and evaluated feasible loading conditions under normal, off-normal, and accident conditions. Two types of metal and concrete storage casks were used in the evaluation. Structural integrity was assessed by comparing load combinations and stress intensity limits under each condition. Evaluation results showed that the storage cask had secured structural integrity as it satisfied the stress intensity limit under normal, off-normal, and accident conditions. These results can be used as baseline data for the detailed design of high burnup storage casks.

To non-destructively determine the burnup of a spent nuclear fuel assembly, it is essential to analyze the nuclear isotopes present in the assembly and detect the neutrons and gamma rays emitted from these isotopes. Specifically, gamma-ray measurement methods can utilize a single radiation measurement value of 137Cs or measure based on the energy peak ratio of Cs isotopes such as 134Cs/137Cs and 154Eu/137Cs. In this study, we validated the extent to which the results of gamma-ray measurements using cadmium zinc telluride (CZT) sensors based on 137Cs could be accurately simulated by implementing identical conditions on MCNP. To simulate measurement scenarios using a lead collimator, we propose equations that represent radiation behavior that reaches the detector by assuming “Direct hit” and “Penetration with attenuation” situations. The results obtained from MCNP confirmed an increase in measurement efficiency by 0.47 times when using the CZT detector, demonstrating the efficacy of the measurement system.

The 300 concrete silo systems installed and operated at the site of Wolsong nuclear power plant (NPP) have been storing CANDU spent nuclear fuel (SNF) under dry conditions since 1992. The dry storage system must be operated safely until SNF is delivered to an interim storage facility or final repository located outside the NPP in accordance with the SNF management policy of the country. The silo dry storage system consists of a concrete structure, liner steel plate in the inner cavity, and fuel basket. Because the components of the silo system are exposed to high energy radiation owing to the high radioactivity of SNF inside, the effects of irradiation during long-term storage must be analyzed. To this end, material specimens of each component were manufactured and subjected to irradiation and strength tests, and mechanical characteristics before and after irradiation were examined. Notably, the mechanical characteristics of the main components of the silo system were affected by irradiation during the storage of spent fuel. The test results will be used to evaluate the long-term behavior of silo systems in the future.