Numerical modeling and scenario composition are needed to characterize the geological environment of the disposal site and analyze the long-term evolution of natural barriers. In this study, processes and features of the hydro-mechanical behavior of natural barriers were categorized and represented using the interrelation matrix proposed by SKB and Posiva. A hydro-mechanical coupled model was evaluated for analyzing stress field changes and fracture zone re-activation. The processes corresponding to long-term evolution and the hydro-mechanical mechanisms that may accompany critical processes were identified. Consequently, practical numerical methods could be considered for these geological engineering issues. A case study using a numerical method for the stability analysis of an underground disposal system was performed. Critical stress distribution regime problems were analyzed numerically by considering the strata’s movement. Another case focused on the equivalent continuum domain composition under the upscaling process in fractured rocks. Numerical methods and case studies were reviewed, confirming that an appropriate and optimized modeling technique is essential for studying the stress state and geological history of the Korean Peninsula. Considering the environments of potential disposal sites in Korea, selecting the optimal application method that effectively simulates fractured rocks should be prioritized.

The solubility and species distribution of radionuclides in groundwater are essential data for the safety assessment of deep underground spent nuclear fuel (SNF) disposal systems. Americium is a major radionuclide responsible for the long-term radiotoxicity of SNF. In this study, the solubility of americium compounds was evaluated in synthetic groundwater (Syn- DB3), simulating groundwater from the DB3 site of the KAERI Underground Research Tunnel. Geochemical modeling was performed using the ThermoChimie_11a thermochemical database. Concentration of dissolved Am(III) in Syn-DB3 in the pH range of 6.4–10.5 was experimentally measured under over-saturation conditions by liquid scintillation counting over 70 d. The absorption spectra recorded for the same period suggest that Am(III) colloidal particles formed initially followed by rapid precipitation within 2 d. In the pH range of 7.5–10.5, the concentration of dissolved Am(III) converged to approximately 2×10−7 M over 70 d, which is comparable to that of the amorphous AmCO3OH(am) according to the modeling results. As the samples were aged for 70 d, a slow equilibrium process occurred between the solid and solution phases. There was no indication of transformation of the amorphous phase into the crystalline phase during the observation period.

최근, 구조설계 기준 및 평가방법의 전문화로 인하여, 선급 규칙의 통합화가 이뤄졌었다. 그 좋은 일례가 국제공통규칙(CSR, Common Structural Rule)이다. 그러나, 종강도 하중이 크게 작용하는 화물창 구역에만 국한하여 세부규정이 제시되어 있고, 선수와 선미부 구조에는 별다른 평가 지침이 없다. 언급한 구역의 구조설계는 조선사의 설계 경험에 의존하여 진행하고 있으며, 선급에서도 명확한 기준 이 없으므로 구조 손상 문제가 발생하더라도 근본적인 원인을 파악하기가 힘들다. 본 연구에서는 선미부에 주로 발생하고 있는 좌굴 손 상의 대표적인 사례에 대한 근본적인 원인을 파악하기 위한 엔지니어링 기반의 해법을 제시하였다. 유한요소해석 모델링 기반 구조 강도 검증을 위하여, 하중 조건, 경계조건, 모델링 방법 그리고 평가 기준에 대한 합리적인 해법을 제시하였다. 선미부에 작용하는 휨 모멘트에 의하여 높이 방향으로 압축하중에 의해서 좌굴이 발생할 가능성이 있으며, 좌굴 강성 증가를 위하여 판 두께 증가 혹은 수직 보강재의 추 가가 필요하다. 앞으로도 이 결과는 유사 운반선의 선미부 구조 강도 검토 시 도움을 줄 것으로 기대된다.

Controller modeling is essential for the design. It allows various control techniques to be simulated in advance, and various interpretations can be performed. If this is not the case, we need to reverse engineering in the real system developed by others. In this paper, controller modeling was reversely designed using the frequency test results of the target system. First, the characteristic equation of the target equipment was based on and a block diagram was assumed. Thereafter, controller variables were estimated using the frequency test results for each of the four control loops. In addition, time response simulations were performed using the estimated controller modeling. This method is thought to be of great help to reverse engineering in situations where there is completed equipment but no controller modeling.

본 연구는 정확한 열환경 평가를 위해 옥외 공간에서 체감하는 열쾌적성과 미기후 모델링을 통해 산출된 열쾌적지수의 차이점을 확인하는 것을 목적으로 한다. 이를 위해 대구광역시에 위치한 대학 캠퍼스 두 곳을 대상으로 하여 열쾌적성을 평가하는 두 가지 방법론을 적용하고 분석 결과를 비교하였다. 첫 번째 방법은 현장 설문조사를 기반으로 하여 시민들의 열쾌적성 정보를 수집하는 것이며, 두 번째 방법은 미기후 모델인 ENVI-met을 활용하여 열쾌적지수(PMV)를 수집하는 것이다. 또한, 열쾌적성에 영향을 미치는 요인을 파악하고자 그늘의 특성과 토지피복 특성을 기준으로 캠퍼스 내 상세 대상지를 선정하고 이러한 요인이 열쾌적성에 미치는 영향을 분석하였다. 분석 결과, 두 대학의 분석 일시와 장소가 달랐으나 방법론별 유사한 양상이 나타났다. 먼저, 설문조사 결과 그늘의 양이 증가할수록 쾌적한 것으로 나타났으며, 토지피복의 종류별 특성과는 다른 결과가 나타났다. 다음으로 모델링 결과 전반적으로 설문조사 결과와 비슷한 양상이 나타나지만 동일한 그늘의 특성을 가지는 세부 대상지에서 토지피복이 열환경에 부정적인 영향을 미치는 피복일 경우 열쾌적지수가 더 높게 나타났다. 결론적으로 설문조사 결과와 미기후 모델링 결과 간에 차이가 존재하며, 정책 반영을 위한 열쾌적성 정보의 수집 시 현장 기반의 체감 더위 정보의 수집을 통해 정확한 정보를 활용할 필요가 있을 것으로 판단된다.

오대산국립공원 내 뱀류 로드킬의 발생 경향 파악 및 예방을 위하여 2006-2017년 사이 공원 내에서 발생한 뱀류 로드킬 자료를 확보 및 분석하였고, 잠재적 발생지 예측을 위하여 종분포모델을 제작하였다. 연구기간 동안 뱀류 로드킬 은 600m 대의 양쪽 환경이 산림-수계인 도로에서 가장 많이 발생하였다. 모델링 결과에서 뱀류 로드킬 발생 가능성은 고도 700m 이하의 하천과의 거리가 25m 부근인 완만한 경사의 도로의 로드킬 발생확률이 높게 나타났다. 국립공원 내 주요 로드킬발생 예측지역은 국도 6호선 도로 위 공원 남쪽 경계로부터 약 2.2㎞ 지역과 약 11.7㎞ 지역이, 지방도 446호선 도로 위 공원 남쪽 경계로부터 약 3.44㎞ 지역이었다. 본 연구결과는 해발고도 700m 이하 수계와 인접한 도로 주변에 우선적으로 대체 일광욕 장소, 생태통로 및 도로의 유입을 막는 울타리의 설치가 산림에서 뱀류 로드킬을 줄이는 효과적인 방안이 될 것을 제시한다.

Medical cyclotrons have been used for dedicated medical of commercial applications such as positron emission tomography (PET) for the past tens of years. These cyclotron facilities have produced positron-emitting radionuclides (i.e. 11C, 13N, 15O, 18F, etc.). Among them, 18F, produced by 18O(p,n)18F reaction is the most widely used which has longer half-life (around 110 m) and lower energy of emitted positrons (around 0.63 MeV). Secondary neutrons produced during 18O(p,n)18F reaction could cause neutron activation of structures, systems, and components of cyclotron facilities. Therefore, International Atomic Energy Agency (IAEA) had addressed that during the operation of cyclotrons, concrete walls become radioactive over time and this radioactivity needs to be characterized for planning of the facility decommissioning. Moreover, several prior studies had estimated the neutron activation and levels of radioactivity of concrete wall of cyclotron facilities. Although those studies assessed the neutron activation of actual cyclotron facilities, however, the purpose of assessment was only for decommissioning each individual facility. Also, the assumptions, conditions or insights of conclusion may be limited to each individual case. For these reasons, this study focused on analysis of effects of major factors (e.g. concrete type, impurity contents of structural materials, etc.) about neutron activation of cyclotron facilities. In this study, the well-known methodology of neutron activation estimation was established and neutron activation products of concrete wall of cyclotron vault was calculated. Also, sensitivity analyses were conducted to figure out the effects of major factors of neutron activation and production of radioactive wastes during decommissioning of the facility. The methodology and results were validated by two steps: comparing with prior studies and comparing with another computer code. Concrete type did not affect that the decision of level of radioactivity waste criteria. Because of relatively longer half-lives, impurity contents of structural materials especially Co and Eu were turned out one of the most important factors for planning the facility decommissioning. It is hard to simply figure out the radioactivity levels of cyclotron facilities, however, rough predictions of minimum period for decay-in-storage as radioactive waste management can be possible with using information of thermal neutron spectra and major impurity nuclides (e.g. 59Co, 151Eu and 153Eu) for minimization of radioactive waste production and relief of charge of radioactive waste management.

During the decommissioning of nuclear facilities, 3D digital model that precisely describes the work environment can expedite the accomplishment of the work. Thus, the workers’ exposure to radiation is minimized and the safety risk to the workers is reduced, while precluding inadvertent effects on the environment. However, it is common that the 3D model does not exist for legacy nuclear facilities as most of the initial design drawings are 2D drawings and even some of the 2D drawings are missing. Even in the case that all of the 2D drawings are intact, these initial design drawings need to be updated using asbuilt data because facilities get modified through years of operation. In those cases, 3D scanning can be a good option to quickly and accurately generate a structure’s actual 3D geometric information. 3D scanning is a technique used to capture the shape of an object in the form of point cloud. Point cloud is a collection of large number of points on the external surfaces of objects measured by 3D scanners. The conversion of point cloud to 3D digital model is a labor-intensive process as a human worker needs to recognize objects in the point cloud and convert the objects into 3D model, even though some of the conversion process can be automated by using commercial software packages. With the aim of full automation of scan-to-3D-model process, deep learning techniques that take point cloud as input and generate corresponding 3D model have been studies recently. This paper introduces an efficient scan simulation method. The simulator generates synthetic point cloud data used to train deep learning models for classifying reactor parts in robotic nuclear decommissioning system. The simulator is built by implementing a ray-casting mechanism using a python library called ‘Pycaster’. In order to improve the speed of simulation, multiprocessing is applied. This paper describes the ray casting simulation mechanism and compares the in-house scan simulator with an open source sensor simulation package called Blensor.

Gyeongju radioactive waste repository has been operated to dispose low and intermediate level radioactive waste in Korea since 2016. Currently, only deep geological disposal facility (1st) is in operation, surface disposal facility (2nd) is scheduled to operate from 2024. As a result, the annual amount of radioactive waste that can be disposed of at deep geological disposal facilities and surface disposal facilities is almost determined. According to this result, it was possible to derive the total annual disposal amount to dispose of all radioactive waste at the Gyeongju repository after landfill disposal facility (3rd) construction. To evaluate it, a predictive model has been designed and radioactive waste generation, storage, and disposal data were input. The predictive model is based on system dynamics, which is useful to analyze the correlation between input variables. As a result of analysis, radioactive waste generation amount and maximum annual radioactive waste disposal were predicted to reach 741,615 drum and 17,030 drum per year respectively. From these results, it seems that the expansion of radioactive waste acceptance system or temporary storage is necessary.

Important medical radionuclides for Positron Emission Tomography (PET) are producing using cyclotrons. There are about 1,200 PET cyclotrons operated in 95 countries based upon IAEA database (2020). Besides, including PET cyclotrons, demands for particle accelerators are continuously increasing. In Korea, about 40 PET cyclotrons are in operating phases (2020). Considering design lifetime (about 30-40 years) and actual operating duration (about 20-30 years) of cyclotrons, there will be demands for decommissioning cyclotron facilities in the near future. PET cyclotron produces radionuclides by irradiating accelerated charged particles to the targets. During this phase, nuclear reactions (18O(p,n)18F etc.) produce secondary neutrons which induce neutron activation of accelerator itself as well as surrounding infrastructures (the ancillary subsystems, peripheral equipment, concrete walls etc.). Generally, experienced cyclotron personnel prefer an unshielded cyclotron because of the repair and maintenance time. In unshielded cyclotron, water cooling systems, air compressor, and other equipment and structures could be existed for operating purposes. Almost all the equipment and structures are consisted of steel, and these affect neutron distribution in vault especially thermal neutron on the concrete wall. In addition, most of them can be classified as very low level radioactive wastes by Nuclear Safety and Security notice (NSSC Notice No. 2020-6). However, few studies were estimating radioactivity concentrations (Bq/g) of surrounding structures using mathematical calculation/simulation codes, and they were not evaluating the effect of surrounding structures on neutron distribution. In this study, by using computational neutron transport code (MCNP 6.2), and source term calculation code (FISPACT- II), we evaluated effect of the interaction between surrounding structures (including surrounding equipment) and secondary neutrons. Discrepancies of activation distribution on/in concrete wall will be occur depending on thickness of structure, distance between structures and walls, and consideration of interaction between structures and neutrons. Throughout this study, we could find that the influence of those structures can affect neutron distribution in concrete walls even if, thickness of the structure was small. For estimating activation distribution in unshielded cyclotron vault more precisely, not only considering cyclotron components and geometry of target, but also, considering surrounding structures will be much more helpful.

The “shadow zone” is defined as a region below a flow obstacle, such as a vault, in unsaturated soils. Due to the capillary discontinuity of the cavity, water saturation on the top and side of the cavity is higher than the ambient saturation. On the bottom of the cavity, however, there is a region where water saturation is lower than ambient saturation. Undoubtedly, a shadow zone may also exist below a LILW disposal vault built in subsurface soils above the water table before the vault is fully degraded. During the degradation, flow in the shadow zone is controlled by the rate of water infiltrating the degrading vault. In this study, as one of the efforts to be made for enhancing safety margin by a realistic safety assessment of the engineered vault type LILW disposal facility, the shadow zone effect is investigated by a numerical parametric study using AMBER code. The conceptual model and data were excerpted from IAEA, ISAM Vault Test Case for the liquid release design scenario. It is assumed that the nearfield barriers degrade with time. In order to compare a visible shadow zone effect, the vault degradation period is assumed to be both 500 and 1,000 years, and the shadow zone depth to be varied according to unsaturated zone lithology. It can be seen that with a shorter shadow zone (2.7 m), radionuclides arrive at the water table earlier than with a full shadow zone (55 m) due to increased advection rate in the unsaturated zone. This effect tends to be more visible in the case of a longer degradation period. For radionuclides with short residence time relative to their half-lives in the unsaturated zone, such as Tc-99 and I-129, the radionuclides are shown to come out because they will arrive sooner, thereby allowing less peak release rate, when the shadow zone effect is considered. Once the vault is completely degraded and the infiltration rate of water flowing through the vault is equal to the ambient rate, the shadow zone effect disappears. In this example calculations using IAEA ISAM Vault Test Case input parameters, it might not be shown a significant shadow zone effect. Nevertheless, when the extent of the shadow zone is determined through more sophisticated hydraulic studies in the unsaturated soils surrounding the vault, the shadow zone effect would be checked up on the realistic near-field radionuclide transport modeling in order to contribute to gaining safety margins for post-closure safety assessment of the Wolsong 2nd phase LILW disposal facility.

Excavation Damaged Zone (EDZ) is created by the excavation of deposition holes and disposal tunnels at high-level radioactive waste repository that causes macro- and micro-fracturing in the surrounding rock. Since EDZ can significantly increase the hydraulic transmissivity in the rock and act as a major pathway of leaked radionuclides, consideration of EDZ in terms of safety assessment is very important. Moreover, long-term stress changes such as stress redistribution due to excavation of nearby deposition holes and disposal tunnels, thermal stress due to temperature rise, effective stress change due to pore pressure change, and swelling pressure of bentonite buffer can increase EDZ size and change in thermal-hydraulic-mechanical properties, and consequently, it can affect the transport of radionuclides. Therefore, in order to analyze the effect of long-term evolution of EDZ on radionuclide transport, it is essential to conduct numerical analysis considering the coupled Thermal-Hydraulic- Mechanical (THM) behavior in EDZ. In order to simulate the behavior of EDZ, coupled THM model was developed using the Adaptive Process-based total system performance assessment framework for a geological disposal system (APro) proposed by the Korea Atomic Energy Research Institute (KAERI). The concept of damage was introduced to demonstrate the jointed rock as a continuous medium. Among several damage models, Mazars damage model was applied in this study. Mazars damage model is the most well-known model for concrete which has similar behavior with rock as brittle material, and the input data of the model can be easily obtained through laboratory testing. If damage occurs due to the influence of thermal-hydraulic-mechanical coupled behavior at the bedrock, the properties change according to the degree of damage, and as a result, the migration of the radionuclide is affected. Based on this conceptual model, radionuclide transport model in the near field considering the long-term evolution of EDZ was developed. To investigate the effect of EDZ in terms of process-based performance assessment, the modeling results with and without EDZ were compared. Finally, by simulating the coupled THM behavior of EDZ with damage model, the effect of long-term evolution of EDZ on radionuclide transport was investigated.

The analysis of uranium migration is crucial for the accurate safety assessment of high-level radioactive waste (HLW) repository. Previous studies showed that the migration of the uranium can be affected by various physical and chemical processes, such as groundwater flow, heat transfer, sorption/ desorption and, precipitation/dissolution. Therefore, a coupled Thermal-Hydrological-Chemical (THC) model is required to accurately simulate the uranium migration near the HLW repository. In this study, COMSOL-PHREEQC coupled model was used to simulate the uranium migration. In the model, groundwater flow, heat transfer, and non-reactive solute transport were calculated by COMSOL, and geo-chemical reaction was calculated by PHREEQC. Sorption was primarily considered as geo-chemical reaction in the model, using the concept of two-site protolysis nonelctrostatic surface complexation and cation exchange (2 SP NE SC/CE). A modified operator splitting method was used to couple the results of COMSOL and PHREEQC. Three benchmarks were done to assess the accuracy of the model: 1) 1D transport and cation exchange model, 2) cesium transport in the column experiment done by Steefel et al. (2002), and 3) the batch sorption experiment done by Fernandes et al. (2012), and Bradbury and Baeyens (2009). Three benchmark results showed reliable matching with results from the previous studies. After the validation, uranium 1D transport simulation on arbitrary porewater condition was conducted. From the results, the evolution of the uranium front with sequentially saturating sites was observed. Due to the limitation of operator splitting method, time step effect was observed, which caused the uranium to sorbed at further sites then it should. For further study, 3 main tasks were proposed. First, precipitation/ dissolution will be added to the reaction part. Second, multiphase flow will be considered instead of single phase Darcy flow. Last, the effect of redox potential will be considered.

This study is to investigate fuel cladding temperature in a transport system for the purpose of developing a methodology for evaluating the thermal performance of spent fuel. Detailed temperature analysis in the transport system is important because the degradation mechanism of the fuel cladding is generally sensitive to temperature and temperature history. In such a system, the magnitude of the temperature change is determined by examining the temperature sensitivity of fuel assemblies and system components including fuel cladding temperature, considering the material properties, component specifications, component aging mechanism, and heat transfer mechanism. The sensitivity analysis is performed using heat transfer models by computational fluid dynamics for the horizontal transport system. The heat transfer within the system by convection, conduction and thermal radiation is calculated by thermal-hydraulic analysis code FLUENT. The calculation region is divided into a basket cell and a transport cask. The thermal analysis of the basket cell is for predicting the fuel cladding temperature. And the reason for analyzing the transport cask is to provide the boundary condition for the basket cell by reflecting the external environmental conditions. Here, the basket cell containing the spent fuel assembly is modeled on the homogeneous effective thermal conductivity. The purpose of this analysis is to evaluate fuel cladding temperatures for the following four main items. That is the effect of surface emissivity changes in basket due to the oxide layer of the fuel cladding, the effect of degradation of the canister backfill helium gas, the effect of fuel assembly position in basket cell on fuel cladding and basket temperatures in canister, and the effect of using the homogeneous effective thermal conductivity model instead of the fuel assembly in basket cell. As a result of the analysis, the maximum temperatures in basket cells are evaluated for the above four items. Thermal margins for each item are investigated for thermal performance requirements (e.g., peak clad temperature below 400oC).