신경내분비종양은 소마토스타틴 수용체의 발현이 증가되어 있다. 소마토스타틴 수용체를 표적으로 하는 소마토스타틴 유사체 옥트레오티드는 오랫동안 신경내분비종양의 기능을 억제하는 치료제로 사용되어 왔다. 옥트레오티드에 핵의학 영상용 방사성동위원소 In-111을 표지하여 환자에 주사한 후 감마카메라로 전신을 촬영하는 기능적 핵의학 영상 또한 오래전부터 사용되었다. 최근에는 옥트레오티드 유사체에 양전자단층촬영(positron emission tomography, PET)용 방사성동위원소를 표지하여 PET/CT를 촬영하게 되었는데 기존 In-111 옥트레오스캔보다 더 선명한 영상을 얻을 수 있다. 나아가 옥트레오티드 유사체에 치료용 방사성핵종을 표지하여 주사하면 신경내분비종양의 전이된 병소를 찾아가서 방사선 치료를 하는 일명 방사선 미사일 치료가 개발되었다. 이는 펩타이드 수용체를 표적으로 하는 핵의학 치료의 일종으로 펩타이드 수용체 방사성핵종 치료(peptide receptor radionuclide therapy, PRRT)라고 한다. 같은 소마토스타틴 수용체 표적 펩타이드를 이용하여 치료 전 기능 영상을 얻어서 PRRT의 대상 환자를 선별할 수 있어 환자 개인맞춤 정밀치료가 가능하다. 또한 Lu-177과 같은 영상용 감마선과 치료용 베타선을 동시에 방출하는 방사성동위원소를 표지하면 치료와 동시에 감마카메라 영상을 얻을 수 있어 주사한 표적치료제의 분포를 매 치료마다 평가할 수 있어 진단과 치료의 합성어인 테라노스틱스가 가능하다.

The aim of this study is to investigate the sorption/ion exchange of radioactive nuclides such as Cs+ and Sr2+ by synthetic Na-micas. In order to prepare Na-micas, two natural micas (phlogopite and biotite) were used as precursor materials. XRD, SEM, and EDS analyses were used to examine material characterization of synthetic Na-micas. Analyses of materials revealed that Na-micas were successfully obtained from natrual micas by K removal treatment. On the other hand, single solute (Cs or Sr) and bi-solute (Cs/Sr) sorption experiments were carried out to determine sorption capacity of Na-micas for Cs and Sr under different pH and ionic strength conditions. Uptake of Cs and Sr by micas in bi-solute system was lower than in single-solute system. Additionally, Langmuir and Langmuir competitive models were applied to describe sorption isotherm of Na-micas. bi-solute system was well described by Langmuir competitive models. For the results obtained in this study, Na-micas could be promising sorbents to treat multi-radioactive species from water and groundwater.

균일한 밀집구조를 지니는 투명한 폴리설폰 필름이 용매를 사용한 표면처리를 통해 개질되었다. 개질은 고분자 필름을 디메털포름아마이드 용매에 1 sec간 담그고 이를 비용매 욕조에 침지시켜 이루어졌다. 개질 전 투명하였던 필름은 개질 후 횐색을 띠며, 많은 기공이 표면에 형성되었다. 비용매로서 물을 사용하였을 때가 이소프로파놀을 사용하였을 때 보다 표면의 불균일도가 증가하였다. 개질된 필름을 사용하여 방사성핵종으로 오염된 지역으로부터 오염물을 채취하였을 때 일반적으로 사용되는 필터페이퍼를 사용하는 것보다 우수한 채취 효율을 보여주었다. 개질된 필름 중에서는 비용매로 물을 사용한 경우 이소프로파놀을 이용한 필름보다 오염물의 채취 효과가 좋은 것으로 나타났다. 한편, 개질된 필름은 고분자 필름의 양쪽 겉표면만을 변화시켜, 필름의 내부는 고유한 밀집도를 유지함으로써, 필터페이퍼 또는 유리섬유를 사용한 오염물 제거 과정에서 나타날 수 있는 매체의 기공을 통한 2차적 오염을 방지하는 효과가 있다.

방사선과의 상호작용에 의하여 섬광이 발생하는 무기형광체(inorganic fluor)인 cerium activated yttrium silicate(CAYS)를 폴리설폰 고분자막에 함침시킴으로써, 형광 용액의 도움 없이 방사능 오염도를 측정할 수 있는 새로운 측정막을 제조하였다. 막의 제조는 두 가지 공정으로 나누어진다. 우선 고분자와 용매의 균일한 1차 제막용액을 유리판 위에 제막 후 용매증발을 통해 폴리설폰 고분자막이 생성되도록 하였다. 고형화된 폴리설폰 필름 표면에 CAYS가 분산된 고분자 제막용액을 2차로 도포시킨 후, 비용매 욕조에 침지시키는 상전환 공정을 이용하거나 용매의 증발을 통한 유리화에 의해 2차 용액의 고형화를 유도함으로써 함침막을 제조하였다. 이렇게 제조된 막의 형상은 치밀한 구조를 지니는 고분자 지지체와 이에 완전히 고착된 CAYS함침막의 이중구조를 지니게되며, 지지체 부분은 막의 안정성을 2차 제막에서 생성된 부분은 기능성의 향상을 이룰 수 있는 구조적 특성을 지닌다. 제조된 함침막에 방사성핵종을 직접 도포하여 방사성핵종의 탐지 특성을 측정하였을 때 효율적인 탐지 특성을 지니는 것으로 확인되었다.

To ensure the safety of disposal facilities for radioactive waste, it is essential to quantitatively evaluate the performance of the waste disposal facilities by using safety assessment models. This paper addresses the development of the safety assessment model for the underground silo of Wolseong Low-and Immediate-Level Waste (LILW) disposal facility in Korea. As the simulated result, the nuclides diffused from the waste were kept inside the silo without the leakage of those while the integrity of the concrete is maintained. After the degradation of concrete, radionuclides migrate in the same direction as the groundwater flow by mainly advection mechanism. The release of radionuclides has a positive linear relationship with a half-life in the range of medium half-life. Additionally, the solidified waste form delays and reduces the migration of radionuclides through the interaction between the nuclides and the solidified medium. Herein, the phenomenon of this delay was implemented with the mass transfer coefficient of the flux node at numerical modeling. The solidification effects, which are delaying and reducing the leakage of nuclides, were maintained the integrity of the nuclides. This effect was decreased by increasing the half-life and the mass transfer coefficient of radionuclides.

When the parent radionuclide decays, the progeny radionuclide is produced. Accordingly, the dose contribution of the progeny radionuclide should be considered when assessing dose. For this purpose, European Commission (EC) and International Atomic Energy Agency (IAEA) provide weighting factors for dose coefficient. However, these weighting factors have a limitation that does not reflect the latest nuclide data. Therefore, in this study, we analyzed the EC and IAEA methodology for derivation of weighting factor and used the latest nuclide data from ICRP 107 to derive weighting factors for dose coefficient. Weighting factor calculation is carried out through 1) selection of nuclide, 2) setting of evaluation period, and 3) derivation based on ICRP 107 radionuclide data. Firstly, in order to derive the weighting factor, we need to select the radionuclides whose dose contribution should be considered. If the half-life of progeny radionuclides sufficiently short compared to the parent radionuclide to achieve radioactive equilibrium, or if the dose coefficient is greater of similar to that of the parent radionuclide and cannot be ignored, the dose contribution of the progeny radionuclide should be considered. In order not to underestimate the dose contribution of progeny radionuclides, the weighting factors for the progeny nuclides are taken as the maximum activity ratio that the respective progeny radionuclides will reach during a time span of 100 years. Finally, the weighting factor can be derived by considering the radioactivity ratio and branch fraction. In order to calculate the weighting factor, decay data such as the half-life of the radionuclide, decay chain, and branch fraction are required. In this study, radionuclide data from ICRP 107 was used. As a result of the evaluation, for most radionuclides, the weighting factors were derived similarly to the existing EC and IAEA weighting factors. However, for some nuclides, the weighting factors were significantly different from EC and IAEA. This is judged to be a difference in the half-life and branch fraction of the radionuclide. For example, in the case of 95Zr, the weighting factor for 95mNb showed a 35.8% difference between this study and previous study. For ICRP 38, when 95Zr decays, the branch fraction for 95mNb is 6.98×10-3. In contrast, for ICRP 107, the branch fraction is 1.08×10-2, a difference of 54.7%. Therefore, the weighting factor for the dose coefficient based on ICRP 107 data may differ from existing studies depending on the half-life and decay information of the nuclide. This suggests the need for a weighting factor based on the latest nuclide data. The results of this study can be used as a basis for the consideration of dose contributions for progeny radionuclides in various dose assessments.

In order to apply indirect methods (such as scaling factors) to assess the radionuclide inventory of waste generated by nuclear power plants, it is essential to first evaluate the correlation coefficient between key radionuclides and those that are difficult to measure (DTM). The benchmark for the correlation coefficient (r) applied in indirect assessments is set at 0.6, and its significance can vary based on both its value and the size of the dataset. For instance, deriving a correlation coefficient using three data points versus utilizing a dataset with a hundred data points would yield different implications. This study addresses the variance in correlation coefficients based on data selection and presents a methodology for validating the significance of these coefficients. Additionally, we will discuss how these variances may impact the results of indirect assessments, such as scaling factor evaluations.

Recently, the nuclear decommissioning and environmental restoration industries has significantly attracted as a new industry field due to the decision to decommission the KORI#1 and WOLSONG #1 nuclear power plant. In order to dispose of the decommissioning radioactive wastes generated during nuclear decommissioning, proper analysis is required, and disposal decisions are determined based on the analysis results. When dismantling a nuclear power plant, a few thousand of tons decommissioning waste are produced, so these require analysis for proper disposal. Therefore, a radionuclide facility for decommissioning waste analysis is essential for the disposal of the large quantities of decommissioning waste generated during nuclear power plant decommissioning. Korea Research Institute of Decommissioning (KRID) was established radionuclide analysis facilities to address above issues and support nuclear power plant decommissioning projects. The plan is to perform classification by type and radionuclide for all waste produced during nuclear power plant decommissioning and to support the disposal of radioactive wastes. In addition, we plan to establish validation methods for samples where verification methods are not established, in order to conduct efficient analysis and management. In this presentation, we will introduce the radionuclide facility currently under construction at KRID and present the space design, equipment layout, and utilization plans.

We conducted safety assessments for the disposal of spent resin mixed waste after the removal of beta radionuclides (3H, 14C) in a landfill facility. The spent resin tank of Wolsong nuclear power plant is generated by 8:1:1 weight ratio of spent ion exchange resin, spent activated carbon and zeolite. Waste in the spent resin tank was classified as intermediate-level radioactive waste due to 14C. Other nuclides such as 60Co and 137Cs exhibit below the low-level radioactive waste criteria. The techniques for separating mixed waste and capturing 14C have been under development, with a particular focus on microwave-based methods to remove beta radionuclides (3H, 14C) from spent activated carbon and spent resin within the mixed waste. The spent resin and activated carbon within the waste mixture exhibits microwave reactivity, heated when exposed to microwaves. This technology serves as a means to remove beta isotopes within the spent resin, particularly by eliminating 14C, allowing it to meet the low-level radioactive waste criteria. Using this method, the waste mixture can meet disposal requirements through free water and 3H removal. These assessments considered the human intrusion scenarios and were carried out using the RESRAD-ONSITE code. The institutional management period after facility closure is set at 300 years, during which accidental exposures resulting from human intrusion into the disposal site are accounted for. The assessment of radiation exposure to intruders in a landfill facility included six human intrusion scenarios, such as the drilling scenario, road construction scenario, post-drilling scenario, and post-construction scenario. Among the six human intrusion scenarios considered, the most conservative assessment about annual radiation exposure was the post-drilling scenario. In this scenario, human intrusion occurs, followed by drilling and residence on the site after the institutional management period. We assumed that some of the vegetables and fruits grown in the area may originate from contaminated regions. Importantly, we confirmed that radiation doses resulting from post-institutional management period human intrusion scenarios remain below 0.1 mSv/y, thus complying with the annual dose limits for the public. This research underscores the importance of effectively managing and securing radioactive waste, with a specific focus on the safety of beta radionuclide-removed waste during long-term disposal, even in the face of potential human intrusion scenarios beyond the institutional management period.

Deep disposal facility for High-Level radioactive Waste (HLW) uses a multi-barrier system to prevent the leakage of radionuclide. As a part of the engineered barrier, bentonite is primarily considered as the main buffering material. This is due to the adsorption and swelling properties of the bentonite, which are expected to effectively impede leakage of the radionuclide. In many cases, adsorption is generally regarded as occurring only within the buffer zone. However, several research has been conducted to explore the possibility of bentonite intrusion into the Excavation- Damaged Zone (EDZ) generated during excavation processes, because of the swelling properties of the bentonite. Generally, for host rock near the deep disposal facility such as granite, groundwater flows through the fracture network. Therefore, analysis of the characteristics of the fracture network is essential for predicting the behavior of radionuclide in groundwater. Accordingly, the bentonite intrusion into the fracture network is critical for safety assessment of the deep disposal facility. To analyze this, hydro-geochemical model was established utilizing COMSOL Multiphysics and PHREEQC, observing changes of the behavior of U (VI) along fracture network due to the swelling of bentonite. Modeling was conducted with progressively changing intrusion depth of the bentonite. According to the results, the behavior of U (VI) exhibited significant changes depending on the connectivity of the fractures. Based on the distribution characteristics of the fracture network, heterogeneous groundwater flow was observed. U (VI) was transported through the preferential pathway, which indicates high connectivity, due to the rapid groundwater flow. Notably, when changing the intrusion depth of bentonite, significant differences in behavior of U (VI) were observed in the 0-20 cm case. In contrast, as the intrusion depth increased, it was observed that differences became less evident. These results indicate that changes in the properties of fracture network in EDZ due to the swelling of bentonite significantly influence the behavior of U (VI).

Safety assessments for geological disposal systems extend over tens of thousands of years, taking into account the radiotoxicity decay period of spent nuclear fuel. During this extensive period, the biosphere experiences multiple glacial cycles, and fluctuations in seawater amounts, attributed to the formation and melting of glaciers, lead to global sea level changes known as eustacy. These sea level changes can directly influence the land-sea interface and groundwater flow dynamics, consequently affecting the pathways of radionuclide transport - an essential element of dose assessment. Therefore, this study aims to investigate how glacial cycles and sea level changes impact radionuclide transport within geological disposal systems, especially in the biosphere. To achieve this objective, we obtained climate evolution data including sea level changes for the Korean Peninsula over a 200,000-years, simulated by a General Circulation Model (GCM). These data were then employed to predict site and hydrology evolutions. The study site was conceptualized biosphere of Artificial Disposal System (ADioS), and we utilized the Soil and Water Assessment Tool (SWAT) to simulate hydrological evolution. These datasets, encompassing climate, site, and hydrology evolution, were collectively employed as inputs for the biosphere module of Adaptive Process-Based Total System Performance Assessment Framework (APro). Subsequently, the APro’s biosphere module calculated radionuclide transport in groundwater flow and its release into surface water bodies, considering the influences of glacial cycles and sea level changes. The results show that hydrologic changes due to sea level change are relatively minor, while the impact of sea level change on groundwater flow and discharge is significant. Additionally, we identified that among the water bodies within ADioS, including rivers, lakes, and oceans, the ocean exhibits the most substantial radionuclide outflow throughout the entire period. The spatiotemporal distributions of radionuclides computed within APro will be further processed into a grid format and used as input for the dose assessment module. Through this study, it was possible to determine the impact of long-term glacial cycles and sea level changes on radionuclide transport. Additionally, this module can serve as a valuable tool for providing the spatiotemporal variability of radionuclides required for enhanced dose assessments.

The increasing accumulation of spent nuclear fuel has raised interest in High-Level Waste (HLW) repositories. For example, Sweden is under construction of the KBS-3 repository. To ensure the safety of such HLW repository, various countries have been developing assessment models. In the Republic of Korea, the Korea Atomic Energy Research Institute has been developing on the AKRS model. However, traditional safety assessment models have not considered the fracture growth in the far-field host rock as a function of time. As repository safety assessments guarantee safety for million years, sustained stress naturally leads to the progressive growth of fractures as time goes on. Therefore, it becomes essential to account for fracture growth in the surrounding host rock. To address this, our study proposes a new coupling scheme between the Fracture growth model and the radionuclide transport model. That coupling scheme consists of the Cubic Law model as a fracture growth function and the GoldSim code which is a commercial software for radionuclide transport calculations. The model that adopting such fracture growth functions showed an increase of up to 15% in the release of radionuclide compared to traditional assessment models. our observations indicated that crack growth as a function of time led to an increase in hydraulic conductivity that allowed more radionuclide transport. Notably, these findings show the significance of adopting fracture growth models as a critical element in evaluating the safety of nuclear waste repositories.

The final disposal of Spent Nuclear Fuel (SNF) will take place in a deep geological repository. The metal canister surrounding the SNF is made of cast iron and copper, designed to provide longterm containment of radionuclides. Canister is intended to be safeguarded by a multiple-barrier disposal system comprising engineered and natural barriers. Colloids and gases are mediators that can accelerate radionuclide migration and influence radionuclide behavior when radionuclides leak from the canister at the end of its service life. It is very important to consider these factors in the assessment of the long-term stability of deep dispoal repository. An experimental setup was designed to observe the acceleration of nuclide behavior due to gas-mediated transport in a simulated environment with specific temperature and pressure conditions, similar to those of a deep disposal repository. In this study, experiments were conducted to simulate gas flow within an engineered barrier under conditions reflecting 1000 years post repository closure. The experiment utilized bentonite WRK with a dry density of 1.61 g/cm³ after compaction. The compacted bentonite was subsequently saturated under a water pressure of 5 MPa, equivalent to the hydrostatic pressure found 500 meters underground. Gas was introduced into the saturated bentonite at different pressures to assess the permeation behavior of the bentonite relative to gas pressure variations. Consequently, it was observed that under specific pressures, gas permeated the saturated bentonite, ascending in the form of bubbles. Furthermore, it was noted that when a continuous flow was initiated within the bentonite, erosion took place, leading to the buoyant transportation of eroded particles upward by the bubbles. The particles transported by the bubbles had a relatively small particle size distribution, and cesium also tended to be transported by the bubbles and moved upward. When high-pressure gas is generated at the interface of the canister and the buffer, flow through the buffer can occur, and cationic nuclides such as cesium and strontium can be attached to the gas bubble and migrate. However, the pressure of the gas to break through the saturated buffer is very high, and the amount of cesium transported by the gas bubbles is very limited.