In the deep geological repository, a considerable quantity of cementitious materials is generally used for structural stability of subcomponents such as grout and concrete plug of disposition tunnel. Strong alkaline leachates (pH>13) are produced after cement is dissolved by groundwater inflow from bedrock. When the leachates are transported to bentonite porewater (e.g. buffer and backfill) and thereby water exchange occurs, the physical properties of bentonite such as swelling capacity and hydraulic conductivity are changed, which eventually affects the safety function and long-term stability of engineered barrier system (EBS). Thus, in this paper, we reviewed the performance assessment methodology for cement-bentonite interaction in the operating license application for the Finnish deep geological repository, and suggested what to prepare for the analysis on the domestic disposal facility. In Finland, thermal-hydraulic-chemical analysis for dissolution of montmorillonite by alkaline leachates resulting from cement degradation during the saturation of bentonite was carried out using PRECIP code. From this analysis, it was confirmed that effect on pH was considered to be more significant than that on temperature and bentonite saturation. As a result of this analysis, it was predicted that all primary minerals (including montmorillonite, quartz, and calcite) were dissolved and some secondary minerals (e.g. chalcedony and celadonite) was precipitated by alkaline cement leachates transported to the bentonite. In addition, it was shown that silica was preferentially released while the montmorillonite was dissolved, thus cementation of the bentonite was occurred. Through this phenomenon, the swelling capacity of bentonite is reduced and the hydraulic conductivity of bentonite is increased, which have a significant impact on the performance of the buffer and backfill. Considering this, study on spreading of alkaline leachates, which is a condition for dissolution of montmorillonite, is necessary for the performance assessment of the domestic deep geological repository. However, this requires the site-specific data for the following in the disposal site: (a) distribution in fractured bedrock and pore structure (e.g. porosity, pore size distribution and pore morphology) in the bedrock, (b) hydraulic gradient and salinity concentration of groundwater, and (c) flux and velocity of groundwater. Results of this study is considered to be directly utilized to the conceptual design and performance assessment of the deep geological repository in Korea, provided that additional data on microbiological properties of groundwater are obtained for the site selected.

The Korean Nuclear Safety and Security Commission has established a general guideline for the disposal of high-level waste, which requires that radiological effects from a disposal facility should not exceed the regulatory safety indicator, a radiological risk. The post-closure safety assessment of the disposal facility aims to evaluate the radiological dose against a representative person, taking into account nuclide transport and exposure pathways and their corresponding probabilities. The biosphere is a critical component of radiation protection in a disposal system, and the biosphere model is concerned with nuclide transport through the surface medium and the doses to human beings due to the contaminated surface environment. In past studies by the Korea Atomic Energy Research Institute (KAERI), the biosphere model was constructed using a representative illustration of surface topographies and groundwater conditions, assuming that the representative surface environment would not change in the future. Each topography was conceptualized as a single compartment, and distributed surface contamination over the geometrical domain was abstracted into 0D. As a result, the existing biosphere model had limitations, such as a lack of quantitative descriptions of various transport and exposure pathways, and an inability to consider the evolution of the surface environment over time. These limitations hinder the accurate evaluation of radiological dose in the safety assessment. To overcome these limitations, recent developments in biosphere modeling have incorporated the nuclide transport process over a 2D or 3D domain, integrating the time-dependent evolution of the surface environment. In this study, we reviewed the methodology for biosphere modeling to assess the radiological dose given by distributed surface contamination over a 2D domain. Based on this review, we discussed the model requirements for a numerical module for biosphere dose assessment that will be implemented in the APro platform, a performance assessment tool being developed by the KAERI. Finally, we proposed a conceptual model for the numerical module of dose assessment.

In the engineered barrier system of deep geological disposal repository, complex physicochemical phenomena occur throughout the entire disposal time, consequently impacting the safety function. The bentonite buffer, a significant component of the engineered barrier system, can be geochemically altered due to the changes in host rock groundwater, temperature, and redox condition. Such changes may have direct or indirect effects on radionuclide migration in case of canister failure. Therefore, a modeling tool that accounts for coupled thermal-hydraulic-mechanical-chemical (THMC) processes is necessary for the safety assessment. To this end, the Korea Atomic Energy Research Institute (KAERI) has developed the APro, a modeling interface for conducting safety assessment of deep geological disposal repository. The APro considers coupled THMC processes that influence radionuclide migration. Here, the solute transport considering thermal and hydraulic processes are calculated using the COMSOL multi-physics, while geochemical reactions are carried out in PHREEQC. The two software are coupled using a sequential non-iterative operator splitting approach, and transport of non-water H, non-water O, and charge were additionally considered to enhance the coupling model stability. Finally, the applicability of APro to simulate long-term geochemical evolution of bentonite was demonstrated through benchmark studies to evaluate the effects of mineral precipitation/dissolution, temperature, redox, and seawater intrusion.

Currently, there are 25 nuclear power plants (NPPs) in operation in Korea, including 22 pressurized water reactors (PWRs) and three pressurized heavy water reactors (PHWRs). Two NPPs, including Kori Unit 1 and Wolsong Unit 1, are permanently shut down and awaiting decommissioning. If Kori Unit 2, which is expected to be permanently shut down soon, is included, the number of decommissioning NPPs will be increased to three. Spent fuels (SFs) are continuously generated during the NPP operation, which are stored in an SF storage pool in NPPs to cool down the decay heat emitted from SFs. For safe NPP operation, SFs must be regarded as waste, and a disposal site must be selected to isolate SFs. However, an appropriate site has yet to be selected in Korea. SFs contain long-lived nuclides with a high specific activity. For disposal, it is important to characterize the nuclides in the fuels and delay the migration of the nuclides to the environment when SFs are placed in a future disposal facility. If the disposal container is broken, the nuclides in the fuels escape from the filling material, such as bentonite. These escaped nuclides are dissolved in groundwater and migrate to the surface of the earth. Thus, it is possible to assess the radiological impact, such as the exposure dose during and after the disposal, if the types and characteristics of nuclides in SFs are known. This study investigated the nuclides in SFs and identified exposure scenarios that may occur in the disposal process of SFs and migration characteristics when the nuclides leak into groundwater to propose a dose assessment methodology for workers and the public.

Two sets of analyses for the cases of groundwater release to well and sea ecosystems were conducted for the environmental impact assessment of high-level radioactive waste disposal facilities. After obtaining the respective BDCF (Biosphere Dose Conversion Factor) results for the scenarios of well-farming and marine water fishing using different biosphere assessment conceptual models implemented in ECOLEGO, they were compared each other. The purposes of these analyses are to identify reference generic biosphere conceptual models and to get insight on model uncertainty. In this study, the endpoint used for the comparison of the ECOLEGO biosphere models was the socalled Biosphere Dose Conversion Factor (BDCF), which is defined as the maximum value of the total dose to the exposed group, in Sv/yr, resulting from a continuous unit release of 1 Bq/yr during the whole simulation time either to the well compartment (BDCF_Well) or to the marine water compartment (BDCF_Sea). The radionuclides considered in the comparison were Cs-137, I-129, Nb-94, Ni-59, Ni- 63, Sr-90 and Tc-99. The conceptual models used in the biosphere assessment of the releases to a well are based on models that have been used by the DOE (simple-soil model) and SKB (complex-soil model) in safety assessments of radioactive waste repositories, respectively. Difference between two conceptual models used in the assessment of the releases to a sea is the number of compartments representing the sea; i.e., one model represents the sea with one compartment for the water and one for the sediment (singlecompartment model), whereas the alternative model uses two compartments for the water and the sediments: one for the inner coast and one for the outer coast (double-compartment model). The results of the BDCF_Well to a farmer obtained with the DOE and SKB models are shown to be very close to each other. Despite the differences in conceptual models and parameters, the results are within a maximum difference of a factor of 4. The results from the SKB model were higher for all radionuclides. The values of the BDCF_Sea obtained with the single- and double-compartment models are shown to be larger differences with a maximum order of 2. For all studied radionuclides, the double-compartment model produces higher BDCFs than does the single-compartment model. The differences would be due to activity concentrations in both water and sediments. Since the hydrodynamic behavior assumed for flow in the sea could significantly influence the dilution volumes and hence the concentrations, it is found that site-specific investigations are necessary to establish an appropriate marine biosphere conceptual model.

IAEA safety standards document and international programs (such as BIOMASS) related to the assessment of the biosphere around High Level Radioactive Waste (including Spent Nuclear Fuel) repositories require the assessment of the biosphere to use the assumption that the current natural environment and human society will be maintained, and at the same time, the evolution of the distant future changes also need to be taken into account. In Korea, which has not designated candidate disposal sites, it is necessary to investigate and predict the current state and future changes of the natural environment throughout Korea and apply it practically to Biosphere assessment (for BDCF derivation) for candidate disposal sites suitability assessment and Safety Case (for performance assessment) preparation for design, construction, operation, and post-closure management. To this end, the natural environment in the fields of Topography, Geology, Soil, Ecology, Weather and Climate, Animals and Plants, Hydrology, Ocean, Land-use, etc. and human society in the fields of Population Distribution, Spatial-Planning, Urban Form, Industrial-Structure, Lifestyle etc. are being investigated in the context of current status, past change records, and future change potential in the Korean Peninsula. This paper summarizes those investigations to date. This study referred Biomass-6 [IAEA] and National Atlas I (2019)/II (2020)/III (2021) [National Geographic Information Institute of the Korea Ministry of Land, Infrastructure and Transport].

Since the first operation of the Gori No. 1 nuclear power plant in Korea was started to operate in 1978, currently 24 nuclear power plants have been being operated, out of which 21 plants are PWR types and the rest are CANDU types. About 30% of total electricity consumed in Korea is from all these nuclear power plants. The accumulated spent nuclear fuels (SNFs) generated from each site are temporarily being stored as wet or dry storage type at each plant site. These SNFs with their high radiotoxicity, heat generating, and long-lived radioactivity are currently the only type of high-level radioactive waste (HLW) in Korea, which urgently requires to be disposed of in deep geological repository. Studies on disposal of HLW in various kind of geological repositories have been carried out in such countries as Sweden, Finland, United States, and etc. with their own management policies in consideration of their situations. In Korea long-term R&D research program for safe management of SNF has also been conducted during last couple of decades since around 1997, during which several various type of disposal concepts for disposal of SNFs in deep geological formations have been investigated and developed. The first concept developed was KAERI Reference Disposal System (KRS) which is actually very much similar to Swedish KBS-3, a famous concept of direct disposal of SNF in stable crystalline rock at a depth of around 500 m which has been regarded as one of the most plausible method worldwide to direct disposal of SNF. The world first Finnish repository will be also this type. Since the characteristics of SNF discharged from domestic nuclear reactors have been changed and improved, and burnup has sometimes increased, a more advanced deep geological repository system has been needed, KRS-HB (KRS with High Burnup SNF) has been developed and in consideration of the dimensions of SNFs and the cooling period at the time point of the disposal time, KRS+, a rather improved disposal concept has also been subsequently developed which is especially focused on the efficient disposal area. Recently research has concentrated on rather advanced disposal technology focused on a safer and more economical repository system in recent view of the rapidly growing amount of accumulated SNF. Especially in Korea the rock mass and the footprint area for the repository extremely limited for disposal site. Some preliminary studies to achieve rather higher efficiency repository concept for disposal of SNF recently have already been emphasized. Among many possible ones for consideration of design for high-efficiency repository system, a double-layered system has been focused which is expected to maximize disposal capacity within the minimum footprint disposal area. Based on such disposal strategy a rather newly designed performance assessment methodology might be required to show long-term safety of the repository. Through the study some prerequisites for such methodological development will be roughly checked and investigated, which covers FEP identification and pathway and scenario analyses as well as preliminary conceptual modeling for the nuclide release and transport in near-field, far-field, and even biosphere in and around the conceptual repository system.

The deep geologic repository (DGR) concept is widely accepted as the most feasible option for the final disposal of spent nuclear fuels. In this concept, a series of engineered and natural barrier systems are combined to safely store spent nuclear fuel and to isolate it from the biosphere for a practically indefinite period of time. Due to the extremely long lifetime of the DGR, the performance of the DGR replies especially on the natural geologic barriers. Assessing the safety of the DGR is thus required to evaluate the impacts of a wide range of geological, hydrogeological, and physicochemical processes including rare geological events as well as present water cycles and deep groundwater flow systems. Due to the time scale and the complexity of the physicochemical processes and geologic media involved, the numerical models used for safety evaluation need to be comprehensive, robust, and efficient. This study describes the development of an accessible, transparent, and extensible integrated hydrologic models (IHM) which can be approved with confidence by the regulators as well as scientific community and thus suitable for current and future safety assessment of the DGR systems. The IHM under development can currently simulate overland flow, groundwater flow, near surface evapotranspiration in a modular manner. The IHM can also be considered as a framework as it can easily accommodate additional processes and requirements for the future as it is necessary. The IHM is capable of handling the atmospheric, land surface, and subsurface processes for simultaneously analyzing the regional groundwater driving force and deep subsurface flow, and repository scale safety features, providing an ultimate basis for seamless safety assessment in the DGR program. The applicability of the IHM to the DGR safety assessment is demonstrated using illustrative examples.

Owing to the increase in saturation rate of the spent fuel storage pond in the Kori nuclear power plant, the interim spent fuel dry storage facility is scheduled to be constructed at the Kori site. To implement safeguards in the new dry storage facility effectively, the concept of “Safeguards-by- Design” (SBD) should be applied to reflect nuclear safeguard provisions in the earliest design stages. Detailed design information pertaining to dry storage facilities has not been determined; however, the design information related to safeguards have been inferred using case studies and interviews with nuclear power plant operators worldwide. On the basis of the results of the case studies on spent fuel dry storage facilities for light water reactors, most countries apply the metal cask method in containment buildings considering safety. Furthermore, Korean operators are also considering the same method owing to tight licensing schedules and safety issues. Using the Facility Safeguardability Assessment (FSA) methodology (one of the safeguard evaluation methodologies), the difference in design between the heavy water reactor spent fuel dry storage facility, an established IAEA safeguards approach reference nuclear facility, and the light water reactor spent fuel dry storage facility (the new nuclear facility) were analyzed. Two major differences were noted as issues pertaining to potential safeguards. First, the difference in design and transport method in terms of the difference in size and weight of the spent nuclear fuel is important; light water reactor fuel is 20 times heavier than heavy water reactor that needs partial defect inspection in assemblies. Second, the difference in safeguard approach owing to the difference between the modular storage method in heavy water reactor and the container type storage method in light water reactor must be considered; movable storage cask renders the IAEA surveillance approach difficult. The results of this study can be used to identify the safeguards requirements in advance, enabling the operator to design new dry storage facilities resulting in timely and cost-effective implementation.

신기술에 대한 투자규모가 크게 증가하고 있고, 과학기술이 사회 전반적으로 대규모·복합적인 영향을 미치고 있다. 정부는 새로운 신기술이 사회에 수용되려면 과학기술에 대한 국민적 관심이 큰 만큼 기술이 국민과 사회에 미치는 영향에 대해서 잘 평가할 의무가 있다. 이를 위한 사회적 합의를 이루기 위해 기술영향평가를 한다. 이를 위한 방법에 대한 연구가 50년째 지속되고 있고, 다양한 학술연구와 수많은 신기술 대응 정책을 통해 논의 된 이후 시대적 변화와 흐름이 있었기에 글로벌 학계와 정책에서 신기술의 미래사회 영향력에 대한 다각적 분석 등이 강조되고 있다. 본 연구는 선진국의 기술영향평가 창안 방식과 우리나라에서 발전된 방식이 다름에도 불구 하고 그동안 신기술 기술영향평가 고려되어야 하는 예측요소와 단계 등 연구 변화에 주목하였다. 연구결과 인식도에 대해 기술 이해도, 전문성, 성별특성 세 개 요소를 도출하고, 기존 요소에 추가해 신기술 예측요소로 제안하였다. 연구의 결과는 우리나라 과학기술기본법에 의거 개선된 기술영향평가를 실시하는데 있어 학술 및 정책적 뒷받침하는 근거를 제시하였다.

This study aims to present a method to evaluate the relative risk of failure due to liquefaction of domestic small to medium-sized earthfill dams with a height of less than 15 m, which has little information on geotechnical properties. Based on the results of previous researches, a series of methods and procedures for estimating the probability of dam failure due to liquefaction, which calculates the probability of liquefaction occurrence of the dam body, the amount of settlement at the dam crest according to the estimation of the residual strength of the dam after liquefaction, the overtopping depth determined from the amount of settlement at the dam crest, and the probability of failure of the dam due to overtopping was explicitly presented. To this end, representative properties essential for estimating the probability of failure due to the liquefaction of small to medium-sized earthfill dams were presented. Since it is almost impossible to directly determine these representative properties for each of the target dams because it is almost impossible to obtain geotechnical property information, they were estimated and determined from the results of field and laboratory tests conducted on existing small to medium-sized earthfill dams in previous researches. The method and procedure presented in this study were applied to 12 earthfill dams on a trial basis, and the liquefaction failure probability was calculated. The analysis of the calculation results confirmed that the representative properties were reasonable and that the overall evaluation procedure and method were effective.

선박 연료로서 LPG는 현재의 기술과 경제성 등을 고려하였을 때 매력적인 연료이다. 하지만, 아직 LPG 연료 선박의 안전 지침 을 개발 중에 있고, 국내에서는 중소형 선박에 LPG 추진 시스템을 적용한 사례가 없다. 본 연구에서는 국내 최초 개발된 해상용 LPG 엔 진 시스템에 대해 보다 객관적인 위험성 평가를 수행하고 안전 운용 기준을 제안하고자 한다. 우선, 위험과 운전 분석 기법을 통해 동 엔 진 시스템을 5개의 검토 구간으로 분할하고 총 58가지의 위험요소를 식별하였다. 그다음 정성적 평가인 HAZOP 기법의 주관성을 보완하 기 위해 퍼지 이론을 사용하고 검출도, 민감도 등 위험 요인을 추가하여 퍼지 분석적 계층 과정을 통해 위험 요인의 상대적 가중치를 비 교하였다. 그 결과, 5가지의 위험 요인 중, 위험성에 가장 큰 영향을 미치는 위험 요인은 발생 빈도와 심각도로 평가되었다. 마지막으로, 위험 요인에 대한 가중치를 고려하여 위험 순위를 세밀하게 선정하기 위해 퍼지 TOPSIS 기법을 적용하였다. 그 결과, 위험 등급은 47개 그룹으로 구분할 수 있었고, 동 엔진 시스템의 운용 중 가장 위험도가 높은 위험요소는 LPG 공급 라인 유지 보수 중 가스 누출로 분석되 었다. 본 연구에 제안된 기법을 LPG 공급계통 등 다양한 설비에도 적용하여, 향후 LPG 추진 선박의 안전 기준 마련을 위한 위험성 평가 의 표준절차로 활용할 수 있기를 기대한다.

해양산업시설의 위험유해물질 배출이 미치는 사회영향을 평가하고, 기술근거배출허용기준 설정과정에서 요구되는 사회영향평 가 항목 및 방법적용에 대한 시사점을 도출하였다. 연구범위는 인천광역시에 있는 해양산업시설을 대상으로 하였다. 분석결과는 다음과 같다. 첫째, 기업 및 산업에 대한 영향은 ‘큰’ 것으로 평가되었지만, 대기업은 영업이익손실과 도산가능성, 그리고 고용손실의 영향이 적 었고, 소기업과 중기업은 영향이 큰 것으로 나타났다. 둘째, 지역사회 및 경제에 대한 영향은 ‘적은’ 것으로 평가되었으나, 직접적으로 인 과관계를 지니는 어업생산자나 지역관광종사자, 그리고 해당 지역은 그 영향이 집중되어 크게 나타날 수 있기 때문에 이에 대한 세심한 정책적 개입이 필요한 것으로 판단된다. 셋째, 기술근거배출허용기준의 설정방법에서 사회적 손실비용 및 사회적 편익산정 항목과 방법 이 유용하게 적용될 수 있는 것으로 판단된다.