Copper, mainly used as a material for outer canister, generates various corrosion products under aerobic and anaerobic conditions in the operational and/or post-closure phases of the deep geological repository. These products could affect performance of engineering barrier system (EBS) through interaction with surrounding bentonite that makes up the buffer and backfill materials. Accordingly, in this study, we suggested research items to be conducted to minimize degradation of EBS due to copper corrosion products, based on the phenomenological review results for copper corrosion mechanisms and interaction between resultant product and bentonite in the deep geological disposal environment. During the post-closure phase, condition in the disposal facility changes form aerobic to anaerobic over time, and thereby, causes and products of copper corrosion vary. Under aerobic condition, copper corrosion is mainly induced by oxygen (O2) in the repository, chloride (Cl-) and carbonate (CO3 2-) ions from groundwater flowing into the facility, resulting in corrosion products such as cuprite (Cu2O), tenorite (CuO), atacamite (CuCl2·3Cu(OH)2) and malachite (Cu2CO3(OH)2). And, copper corrosion under anaerobic condition is primarily due to hydrogen sulfide (H2S) and sulfate (SO4 2-) in groundwater flowing into the facility, leading to formation of chalcocite (Cu2S) and covellite (CuS) as corrosion products. Depending on environment of the disposal facility, copper corrosion products are dissolved and ionized to Cu2+ in groundwater, and subsequently adsorbed on the nearby smectite. Then, it causes a cation exchange reaction with exchangeable cations in the interlayer of smectite. As a result of reviewing the previous experiments, it was confirmed that Cu2+-exchanged bentonite has a slightly reduced basal spacing and swelling capacity. From the results as above, there is a possibility that performance of EBS may be degraded due to copper corrosion products. To minimize its effect of degradation in the domestic facility, items to be further studied are as follows: (a) Method for reducing copper corrosion such as selection of appropriate material and structure for the canister, and (b) How to control dissolution of copper canister product into groundwater through predicting type and ionization process. The results of this study could be directly used to developing design concept of EBS for the domestic disposal facility and to establishing roadmap of future R&D programs.

In 2012, POSIVA selected a bentonite-based (montmorillonite) block/pellet as the backfilling solution for the deposition tunnel in the application for a construction license for the deep geological repository of high-level radioactive waste in Finland. However, in the license application (i.e. SC-OLA) for the operation submitted to the Finnish Government in 2021, the design for backfilling was changed to a granular mixture consisting of bentonite (smectite) pellets crushed to various sizes, based on NAGRA’s buffer solution. In this study, as part of the preliminary design of the deep geological repository system in Korea, we reviewed history and its rationale for the design change of Finland’s deposition tunnel backfilling solution. After the construction license was granted by the Finnish Government in 2015, POSIVA conducted various lab- and full-scale in-situ tests to evaluate the producibility and performance of two design alternatives (i.e. block/pellet type and granular type) for backfilling. Principal demonstration tests and their results are summarized as follows: (a) Manufacturing of blocks using three types of materials (Friedland, IBeco RWC, and MX-80): Cracking and jointing under higher pressing loads were found. Despite adjusting the pressing process, similar phenomena were observed. (b) 1:6 scale experiment: Confirmation of density difference inhomogeneity due to the swelling of block/pellet backfill and void filling due to swelling behavior into the mass loss area of block/pellet. (c) FISST (Full-Scale In situ system Test): Identification of technical unfeasibility due to the inefficient (too manual) installation process of blocks/pellets and development of an efficient granular in-situ backfilling solution to resolve the disadvantage. (d) LUCOEX-FE (Large Underground Concept Experiments – Full-scale Emplacement) experiment: Confirmation of dense/homogeneous constructability and performance of granular backfilling solution. In conclusion, the simplified granular backfill system is more feasible compared to the block/ pellet system from the perspective of handling, production, installation, performance, and quality control. It is presumed that various experimental and engineering researches should be preceded reflecting specific disposal conditions even though these results are expected to be applied as key data and/or insights for selecting the backfilling solution in the domestic deep geological repository.

As increasing number of stray cats, it recently has more chance for stray cats to invade people area. Inevitably, the number of cases of inconvenience and damage caused by street cats is also increasing proportionally. Therefore, we investigated and analyzed the current status and implementation of TNR in Korea and suggested the direction in which Korea's policy should go. Domestic TNR has been implemented for about 20 years, but the ratio of stray cats to which TNR is applied insufficient and post-management after TNR is also insufficient. In the case of other countries, a large percentage of TNR is conducted by selecting TNR target cats based on scientific grounds and adoption or additional observation is carried out after TNR. Based on this strategy, the conditions for successful TNR need to expand the decision-making authority of veterinarians involved in TNR and actively invigorate subsequent solutions such as adoption. In addition, the policy should be improved in the way of expanding management and supervision after TNR through food service centers and shelters.

In this study, a survey focusing on the status of clothing interest, inconveniences resulting from clothing, preferred design items, etc. was conducted on 364 elderly women to suggest aesthetically and functionally appropriate indoor wear design for at home elderly women aged 60 years or older. The survey results showed that in general, the respondents’ interest in clothing was high, and more respondents in their 70s or older had difficulty in the action of opening and closing. With respect to considerations when purchasing clothes, color was considered more important than design as respondent’s age increased, and size was regarded as the most important factor especially among those in their 80s. The preferred top styles were T-shirts and blouses among those in their 60s and 70s, and T-shirts and shirts among those in their 80s. The preferred sleeve lengths were “below the elbow” and “above the wrist” in all age groups. The preferred sleeve hem type was “tightening” in all age groups. The most preferred bottom styles were “straight-leg pants” and “elastic waistband.” This study suggests the design items of indoor wear, including top, bottom, and overgarment for warmth, appropriate for elderly women at home based on the survey results. The study results are expected to serve as basic data necessary for the revitalization of the clothing industry for elderly women.

대한민국 사회가 ‘통일’을 이루기 위해서는 남들보다 북한이탈주민이 대한민국 사회에서 국가 공동체의 한 일원으로서 살아갈 수 있도록 각 분야의 관심과 지원이 필요하다. 특히 북한이탈주민은 북한의 사회주의 경제 체제에서 장기간 노출되면서 대한민국의 자본주의 경제 체제에 적합한 행동 양식을 곧바로 발휘하기 힘드므로 이들의 눈높이에 맞춘 경제 및 금융교육이 필요하다. 대한민국 정부는 자본주의 체제에 적응하는 북한이탈주민을 위한 재사회화 과정에서 다양한 금융교육 자료를 제작 및 활용하고 있다. 그래서 본 연구에서는 금융감독원이 통일부와 함께 발간한 책 <똘똘이와 삼녀의 금융생활 정착기>의 2판을 분석 하면서, 해당 교재가 본래의 출간 목표를 충실히 지키고 있는지 본 연구를 통해 검증하고자 한다. 해당 도서의 자료가 2020년에 새롭게 제정된 금융교육 표준안의 내용을 어떻게 담 고 있는지 단원 별 비중과 내용, 서술 방식 등을 살펴보고, 북한이탈주민의 원활한 금융생 활을 돕기 위해서 해당 교재가 어떠한 방향으로 변화가 필요한지 제안하고자 한다.

The Wolseong Wooden moat is a unique example. Which is a vertical wall made of wood. It shows a changing point how to make the wall by digging a hole and stacking stones vertically. This study tried to to make a assumption about the construction process of the wooden structure found in the Wolseong pit moat. I sorted out wooden elements and then analyzing these features and compared with the results of the excavation. After I made 3D modelling in the order to it was made. This moat is not only a function of digging up the ground to trap water, but also a technique of building structures to maintain walls. It is a valuable material that can show the woodworking engineering techniques of the Silla Dynasty.

As a result of various generation, transmutation, and decay schemes, a wide variety of radionuclides exist in the reactor prior to accident occurrence. Considering all of the radionuclides as the accident source term in an offsite consequence analysis will inevitably take up excessive computer resources and time. Calculation time can be reduced with minimal impact on the accuracy of the results by considering only the nuclides that have a significant effect on the calculation among the potential radioactive sources that may be released into the environment. In earlier studies related to offsite consequence analysis, it is shown that the principal criteria for the radionuclide screening applied are as follows; radionuclide inventory in the reactor, radioactive half-life, radionuclide release fraction to the environment, relative dose contribution of nuclides within a specific group, and radiobiological importance. As a result, it is confirmed that 54, 60, and 69 nuclides are applied to the risk assessment performed in WASH-1400, NUREG-1150, and SOARCA (State-of-the-Art Reactor Consequence Analyses) project in the United States, respectively. In addition, in this study, the technical consultations with domestic and foreign experts were carried out to confirm details on criteria and process for screening out radionuclides in offsite consequence analysis. In this paper, based on the literature survey and technical consulting, we derived the screening process of selecting a list of radionuclides to be considered in the offsite consequence analysis. The first step is to eliminate radionuclides with little core inventory (less than specific threshold) or very short half-lives. However, important decay products of radionuclides that have short half-lives should not be excluded by this process. The next step is to further eliminate radionuclides by considering contribution to offsite impact, which is defined as a product of radioactivity released to the environment (i.e. ‘inventory in the reactor’ times ‘release fraction to offsite’) and comprehensive dose (or risk) coefficient taking into account all exposure pathways to be included. The final step is to delete isotopes that contribute less than certain threshold to any important dose metric through additional computer runs for each important source term. Even though it is presumed that this process is applicable to existing light water reactors and the set of accidents that would be considered in PSA, some of the assumptions or specific recommendations may need to be reconsidered for other reactor types or set of accident categories.

Currently, KHNP-CRI has developed 100 kW plasma torch melting facility to reduce the amount of radioactive waste in nuclear power plant. Plasma torch melting technology uses electric arc phenomena like lightning to melt the target material at a high temperature of about 1,600°C. The technology is applicable to treatment for various types of waste such as combustible, non-combustible and mixed wastes. The volume reduction ratio by the technology is respectively expected to be about 1/60 of combustible wastes and about 1/5 for non-combustible wastes. It is important to discharge the melt without problems in the melting technology. In general, molten slag has properties such as high viscosity and quick solidification. Because of the properties, when discharging into slag container, the final product is accumulated like a mountain. To improve this problem, there is three suggestions; 1) rotation of the slag container, 2) vibration of the slag container, and 3) heating of the slag container.

Bentonite, a material mainly used in buffer and backfill of the engineering barrier system (EBS) that makes up the deep geological repository, is a porous material, thus porewater could be contained in it. The porewater components will be changed through ‘water exchange’ with groundwater as time passes after emplacement of subsystems containing bentonite in the repository. ‘Water exchange’ is a phenomenon in which porewater and groundwater components are exchanged in the process of groundwater inflow into bentonite, which affects swelling property and radionuclide sorption of bentonite. Therefore, it is necessary to assess conformity with the performance target and safety function for bentonite. Accordingly, we reviewed how to handle the ‘water exchange’ phenomenon in the performance assessment conducted as part of the operating license application for the deep geological repository in Finland, and suggested studies and/or data required for the performance assessment of the domestic disposal facility on the basis of the results. In the previous assessment in Finland, after dividing the disposal site into a number of areas, reference and bounding groundwaters were defined considering various parameters by depth and climate change (i.e. phase). Subsequently, after defining reference and bounding porewaters in consideration of water exchange with porewater for each groundwater type, the swelling and radionuclides sorption of bentonite were assessed through analyzing components of the reference porewater. From the Finnish case, it is confirmed that the following are important from the perspective of water exchange: (a) definition of reference porewater, and (b) variations in cation concentration and cation exchange capacity (CEC) in porewater. For applying items above to the domestic disposal facility, the site-specific parameters should be reflected for the following: structure of the bedrock, groundwater composition, and initial components of bentonite selected. In addition, studies on the following should be required for identifying properties of the domestic disposal site: (1) variations in groundwater composition by subsurface depth, (2) variations in groundwater properties by time frame, and (3) investigation on the bedrock structure, and (4) survey on initial composition of porewater in selected bentonite The results of this study are presumed to be directly applied to the design and performance assessment for buffer and backfill materials, which are important components that make up the domestic disposal facility, given the site-specific data.

In buffer, a main component of engineering barrier system (EBS) in the deep geological repository, mass loss is mainly caused by upheave and mechanical erosion. The former is a phenomenon that bentonite in the upper part of the buffer moves to the backfill region due to groundwater intake and swelling. And, the latter is a phenomenon that bentonite on the surface of the buffer moves to the backfill region due to groundwater flow at the interface with host rock as the buffer saturates. Buffer mass loss adversely affects the fulfilment of the safety function of the buffer that is to limit and retard radionuclide release in the event of canister failure. Accordingly, in this paper, we reviewed how to consider this phenomenon in the performance assessment for the operating license application in Finland, and tentatively summarized data required to conduct the analysis for the domestic facility based on the review results. Regarding buffer mass loss, the previous studies carried out in Finland are categorized as follows: 1) experiment on the amount of buffer upheave with groundwater inflow rate (before backfilling), 2) analysis for the amount of buffer upheave with groundwater inflow rate (after backfilling), 3) analysis of buffer erosion rate with groundwater inflow rate, 4) analysis for distribution of the groundwater inflow rate into the buffer for all deposition holes (using ConnectFlow modeling results), and 5) analysis of buffer mass loss with groundwater salinity. Finally, the buffer mass loss distribution table was derived from the results of 1) through 3) by combining with that of 4). Given these studies, the following will be required for the performance assessment for buffer mass loss in the domestic disposal facility: a) distribution table of buffer mass loss for combined interactions taking into account effect of 5) (i.e. 1), 2), 3), and 5) + 4)), and b) Threshold for buffer mass loss starting to negatively affect the fulfilment of the safety function of the buffer. Even though it is judged that the results of this study could be directly applied to developing the design concept of EBS and to conducting the performance assessment in the domestic disposal facility, it is essential to prepare a set of input data reflecting the site-specific design features (e.g. dimension, material used, site, etc.), which include saturation time and groundwater salinity.

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.

Licensees are required to protect critical digital assets (CDAs) in nuclear facilities against cyber-attacks, up to and including design basis threat (DBT), according to「ACT ON PHYSICAL PROTECTION AND RADIOLOGICAL EMERGENCY」. However, CDAs may be excluded from cyber security regulations at nuclear power plant decommissioning, and this may lead to severe consequences if the excluded CDAs contain sensitive information such as the number and location of nuclear fuels and information on security officers. In that case, that information could be leaked to the adversary without adequately removing the information before discarding the CDAs. It can be potentially abused to threaten nuclear facilities inducing radiological sabotage and nuclear material theft. So, controls of sensitive information are needed. This study aims to derive regulatory improvements related to discarding CDAs that have sensitive information by analyzing foreign cases such as IAEA and U.S. NRC. The sensitive information in the IAEA guide is the following: (1) details of physical protection systems and any other security measures in place for nuclear material, other radioactive material, associated facilities, and activities; (2) information relating to the quantity and form of nuclear material or other radioactive material in use or storage; (3) information relating to the quantity and form of nuclear material or other radioactive material in transport; (4) details of computer systems; (5) contingency and response plans for nuclear security events; (6) personal information; (7) threat assessments and security alerting information; (8) details of sensitive technology; (9) details of vulnerabilities or weaknesses that relate to the above topics; (10) historical information on any of the above topics. In the case of the U.S. NRC, they categorize sensitive information into three groups: (1) classified information, (2) safeguard information (SGI), (3) sensitive unclassified non-safeguards information (SUNSI). Classified information is information whose compromise would cause damage to national security or assist in manufacturing nuclear weapons. The SGI concerns the physical protection of operating power reactors, spent fuel shipments, strategic special nuclear material, or other radioactive material. Finally, SUNSI is generally not publicly available information such as personnel privacy, attorney-client privilege, and a confidential source. IAEA recommends protecting the above sensitive information in accordance with NSS No.23-G (Security of Nuclear Information), and NRC protects classified information, SGI, and SUNSI under relative laws. In the case of ROK, if security control measures are enhanced CDAs that possess sensitive information, the risk of information leakage will be decreased when those CDAs are discarded.

When decommissioning nuclear power plant (NPP), the first task performed is cost estimation. This is an important task in terms of securing adequate decommissioning funds and managing the schedule. Therefore, many countries and institutions are conducting continuous research and also developing and using many programs for cost estimation. However, the cost estimated for decommissioning an NPP typically differs from the actual cost incurred in its decommissioning. This is caused by insufficient experience in decommissioning NPPs or lack of decommissioning cost data. This uncertainty in cost estimation can be in general compensated for by applying a contingency. However, reflecting an appropriate standard for the contingency is also difficult. Therefore, in this study, data analysis was conducted based on the contingency guideline suggested by each institution and the actual cost of decommissioning the NPP. Subsequently, TLG Service, Inc.’s process, which recently suggested specific decommissioning costs, was matched with ISDC (International Structure for Decommissioning Costing)’s work breakdown structure (WBS). Based on the matching result, the guideline for applying the contingency for ISDC’s WBS Level 1 were presented. This study will be helpful in cost estimation by applying appropriate contingency guidelines in countries or institutions that have no experience in decommissioning NPPs.

A rod internal pressure increased by fission gas release is major factor that causes degradation during dry storage of spent fuel. Because rod internal pressure is greatly affected by fuel design, operation power history, it is essential to perform complex calculation using performance code to accurately predict rod internal pressure as function of burnup. However, because it is difficult to apply a complex method into dry storage design and to determine rod internal pressure based on conservative way this study presents a simple correlation that can predict an approximate rod internal pressure as function of burnup For the development of simple correlation, rod internal pressure and fuel rod void volume data measured through about 400 PIE (Post Irradiation Examination) data were used. The developed simple correlation can cover various fuel rod arrays, discharged fuel average burnup, operation history, cladding type, burnup range, and information on Westinghouse type fuel rods such as Spain ENUSA, USA EPRI/ANL/ORNL/PNNL, WEC, etc. In this paper, the data of simple correlation determination is briefly introduced, and the data analysis process and results are summarized. Two correlations that can conservatively determine rod internal pressure and free void volume in fuel rod according to fuel rod average burnup were presented, and the effect of initial He fill pressure was evaluated. In particular, the results of Post Irradiation Examination for 46 fuel rods conducted in Korea are also included, so it is expected that newly presented correlations can be used easily in various ways in the domestic research, industry, and academia.

Since the 1992 ‘Joint Declaration of South and North Korea on the Denuclearization of the Korean Peninsula’ was agreed, various negotiations and policies have been conducted. There were policies such as CVID, Strategic Patience, Top-Down Approach, Calibrated Practical Approach, Audacious Initiative and the Geneva Agreement, and 9.19 Joint statement by the six-party talks were signed to denuclearize Korean Peninsula. However, starting with the first nuclear test in 2006, North Korea conducted six nuclear tests to develop atomic bombs, boosted fission bombs, ICBMs, and SLBMs to enhance its weapon capabilities. In addition, nuclear security crisis on the Korean Peninsula has been rising day by day as signs of restarting the North Korea’s Yongbyon 5MWe Graphite-moderated reactor were observed and the possibility of a seventh nuclear test have been increased. Since North Korea’s nuclear issue has a lot of influence on international security, especially on the Northeast Asian countries, a realistic denuclearization policy that reflects North Korea’s current domestic situation along with the international situation is needed. It’s been six months since Russia invaded Ukraine on February 24, 2022. The war between Russia, which has nuclear weapons, and Ukraine, which gave up its possession of nuclear weapons due to the Cooperative Threat Reduction (CTR) program known as the Nunn-Lugar program, is expected to have a significant impact on North Korea, which is considering denuclearization due to UN sanctions on North Korea. Therefore, in this study, based on the war patterns of Russia-Ukraine war, perspective on how it could affect North Korea’s denuclearization is analyzed. Also, significance and limitations of the previous nuclear negotiations, the North Korea’s political regime, the ‘five-year strategy for North Korea’s economic development’ and the ‘five-year plan for North Korea’s economic development’ were analyzed to suggest practical DPRK’s denuclearization policy.

Spent filters with a high radiation dose rate of 2 mSv·hr−1 or more are not easily managed. So far, the Korean policy for spent filter disposal is to store them temporarily at nuclear power plants until the waste filters can be easily managed. Nuclear power plant decommissioning in Korea is starting with Kori unit 1. Volume reduction of waste generated during decommissioning can reduce the cost and optimize the space usage at disposal site. Therefore, efficient volume reduction is a very important factor during the decommissioning process. A conceptual method, based on the experiences of developing 200 and 800 ton compactors at Orion EnC, has been developed considering worker exposure with the followings a crusher (upgrade of compaction efficiency), an automatic dose measuring system with a NaI(Tl) detector, a shield box, an inner drum to prepare for easy handling of drums and packaging, a 30 ton compactor, and an automatic robot system. This system achieves a volume reduction ratio of up to 85.7%; hence, the system can reduce the disposal cost and waste volume. It can be applied to other types of wastes that are not easily managed due to high dose rates and remote control operation necessity.