At the end of 2022 there were 439 nuclear power reactors in operating around the world, including 25 nuclear power reactors of Korea. Domestic nuclear power plants (NPPs) continuously produce low and intermediate-level radioactive waste (LILW) and spent nuclear fuel (SNF). As amount of radioactive waste is increasing and interim storage facilities meet limitation of their capacity, radioactive waste need to be transported. Consequently, the demand for radioactive waste transportation is increasing and the importance of Radiation Risk Assessment Codes (RRACs) for radioactive waste transportation is also on the rise. Considering the domestic situation where all NPPs are located on seaside, the radioactive waste transportation by ship is inevitable and the its risk assessment is very important for safety. Although various researches on radioactive waste transportation risk assessment is being actively conducted, research on domestic radioactive waste maritime transportation is insufficient. In this study, MARINRAD and KM-RAD were used to review on the radioactive waste transportation risk assessment. The result of reviewing shows that MARINRAD used SNF as transporting radioactive materials and ‘SAND87-7067 (1987)’ as nuclide database, whereas KMRAD used LILW and ‘IAEA Technical Report Series-422 (2004)’. To complement these limitations, we present an modernized integrated database by updating data and covering the radioactive materials from LILW to SNF. These results are expected to contribute to the development of RRACs for domestic radioactive waste maritime transportation.

The acceptance criteria for low and intermediate level radioactive waste disposal facilities in Korea to regulate that homogeneous waste, such as concentrated waste and spent resin, should be solidified. In addition, solidification requirements such as compressive strength and leaching test must be satisfied for the solidified radioactive waste solidified sample. It is necessary to develop technologies such as the development of a solidification process for radioactive waste to be solidified and the characteristics of a solidification support. Radioactive waste solidification methods include cement solidification, geopolymer solidification, and vitrification. In general, low-temperature solidification methods such as cement solidification and geopolymer solidification have the advantage of being inexpensive and having simple process equipment. As a high-temperature solidification method, there is typically a vitrification. Glass solidification is generally widely used as a stabilization method for liquid high-level waste, and when applied to low- and intermediate-level radioactive waste, the volume reduction effect due to melting of combustible waste can be obtained. In this study, the advantages and disadvantages of the solidification process technology for radioactive waste and the criteria for accepting the solidified material from domestic and foreign disposal facilities were analyzed.

For the deep geological repository, engineering barrier system (EBS) is installed to restrict a release of radionuclide, groundwater infiltration, and unintentional human intrusion. Bentonite, mainly used as buffer and backfill materials, is composed of smectite and accessory minerals (e.g. salts, silica). During the post-closure phase, accessory minerals of bentonite may be redistributed through dissolution and precipitation due to thermal-hydraulic gradient formed by decay heat of spent nuclear fuel and groundwater inflow. It should be considered important since this cause canister corrosion and bentonite cementation, which consequently affect a performance of EBS. Accordingly, in this study, we first reviewed the analyses for the phenomenon carried out as part of construction permit and/or operating license applications in Sweden and Finland, and then summarized the prerequisite necessary to apply to the domestic disposal facility in the future. In previous studies in Sweden (SKB) and Finland (POSIVA), the accessory mineral alteration for the post-closure period was evaluated using TOUGHREACT, a kind of thermal-hydro-geochemical code. As a result of both analyses, it was found that anhydrite and calcite were precipitated at the canister surface, but the amount of calcite precipitate was insignificant. In addition, it was observed that precipitate of silica was negligible in POSIVA and there was a change in bentonite porosity due to precipitation of salts in SKB. Under the deep disposal conditions, the alteration of accessory minerals may have a meaningful influence on performance of the canister and buffer. However, for the backfill and closure, this is expected to be insignificant in that the thermal-hydraulic gradient inducing the alteration is low. As a result, for the performance assessment of domestic disposal facility, it is confirmed that a study on the alteration of accessory minerals in buffer bentonite is first required. However, in the study, the following data should reflect the domestic-specific characteristics: (a) detailed geometry of canister and buffer, (b) thermal and physical properties of canister, bentonite and host-rock in the disposal site, (c) geochemical parameters of bentonite, (d) initial composition of minerals and porewater in bentonite, (e) groundwater composition, and (f) decay heat of spent nuclear fuel in canister. It is presumed that insights from case studies for the accessory mineral alteration could be directly applied to the design and performance assessment of EBS, provided that input data specific to the domestic disposal facility is prepared for the assessment required.

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.

During the treatment of spent nuclear fuel, radioactive iodine is generated in a liquefied or gaseous form in a specific process. In the case of iodine 129, it is a long-lived nuclide with a very long halflife and has high groundwater mobility under repository conditions. Despite showing a low radioactivity value, research on the management of radioactive iodine from a long-term perspective is continuously being performed. Although research has been conducted using borosilicate glass as a medium for solidifying iodine, compatibility of I in borosilicate glass is very small and the volatility is high in the solidification process. So it is not suitable as a solidified substance of iodine. Therefore, studies on other solidification media to replace them are continuously being conducted. Our research team tried to develop a new medium that can contain iodine in a solidified body stably through a simple heat treatment process and can improve problems such as volatility and waste loading. Iodine is captured as AgI in the Ag ion-exchanged zeolite. So, TeO2, Ag2O, and Bi2O3 having a high AgI loading rate were used as main components. It was named TAB after taking the first letter of each element. In previous studies, the physical properties, structure, and chemical stability of TAB materials were confirmed. PCT (Product Consistent test) was performed to confirm chemical stability. It is mainly used to compare the chemical stability of glass materials with other glass materials, but there are limitations in evaluating the long-term chemical stability of materials. In this experiment, we tried to evaluate the long-term stability of TAB and compare it with borosilicate, which is conventionally used to treat radioactive waste. In addition, we tried to understand the leaching behavior inside the TAB medium. For this purpose, ASTM C1308 test was performed for 365 days, and distilled water and KURT groundwater were used as leachates to examine the effect of ions in the groundwater on the solidified body. To analyze the leaching behavior, ICP-MS and ICP-OES analyses were performed, and the cross-section of the sample after leaching was observed through SEM.

Currently, radioactive waste for disposal has been restricted to low and intermediate level radioactive waste generated during operation of nuclear power plants, and these radioactive wastes were managed and disposed of the 200 L and 320 L of steel drums. However, it is expected that it will be difficult to manage a large amount of decommissioning waste of the Kori unit 1 with the existing drums and transportation containers. Accordingly, the KORAD is currently developing various and largesized containers for packaging, transportation, and disposal of decommissioning waste. In this study, the radiation exposure doses of workers and the public were evaluated using RADTRAN computational analysis code in case of the domestic onroad transportation of new package and transportation containers under development. The results were compared with the domestic annual dose limit. In addition, the sensitivity of the expected exposure dose according to the change in the leakage rate of radionuclides in the waste packaging was evaluated. As a result of the evaluation, it was confirmed that the exposure dose under normal and accident condition was less than the domestic annual exposure dose limit. However, in the case of a number of loading and unloading operations, working systems should be prepared to reduce the exposure of workers.

국내 원자력발전소의 사용후핵연료는 소내저장시설에 보관되어 있으나 저장시설의 용량 확장이 어려움이 있으며, 연구기관 의 연구로에서 발생하는 고준위 방사성폐기물도 자체 보관중이나 영구적으로 저장할 수 없다. 또한 원전의 해체 시에도 고 준위 방사성폐기물이 발생할 것으로 예상된다. 이에 따라, 본 연구에서는 현재 개발된 사용후핵연료 운반용기를 사용하여 고준위 방사성폐기물을 가상의 관리시설로 철도를 통하여 운반하는 경우에 대하여 작업자 및 운반경로 주변 일반인의 예상 피폭선량을 평가하였으며, 그 결과를 국내 법적기준치와 비교하였다. 또한, 고준위 방사성폐기물의 상하차 작업 시 작업자 와 운반용기 간 거리와 운반사고 시 방사성핵종의 누출율의 변화에 따른 피폭선량의 변화에 따른 피폭선량 추이와 운반에 사용되는 열차의 구성에 따른 운반작업자의 피폭선량 변화를 분석하였다. 본 연구에서 설정한 모든 조건에서의 예상피폭선 량은 국내 법적제한치 이하임을 확인하였다.

The adsorption of heavy metals in the waste water carried out on the various domestic clays and waste pottery. The effect of parameters such as pH, temperature, adsorption time and coexisting cations on the adsorption ability and characteristics were investigated to find out whether the clays could be used as adsorbents. Adsorption equilibrium was reached within 20 minutes on all the clays. The optimum pH was found to be above 5. When other cations such as Cu(II) or Zn(II) coexisted with Pb(II), the adsorption amount of Pb(II) decreased because of competing adsorption.

Local government should have waste treatment facilities to provide good service to local residents, even though private recycling is working. There was a problem with plastic waste management in Korea in 2018. Therefore, study was conducted on whether local government has the capacity to handle additional waste streams. The study was conducted, solely using government statistics, on domestic mixed solid waste. The amount of additional plastic waste to be disposed was 2,418 ton/day (incinerate 713 ton/day + landfill 1,705 ton/day), and paper waste was 4,469 ton/day (incinerate 1,104 ton/day + landfill 3,365 ton/day). Current incinerator capacity is sufficient, and if paper waste is added, the incinerator capacity is found to be under1,544 ton/day. Landfill capacity is sufficient even if plastic and paper waste is added, but the residual life of the landfill was reduced from 31.4 years to 25.4 years. Regionally, Gyeongbuk, Daejeon, Jeju, and Sejong should develop new plans for waste management.

2006년 기준 폐기물 처리량 중 매립량은 9,945 천톤, 소각량은 6,940 천톤이었으나 2015년 기준 폐기물 처리량 중 매립량은 13,797 천톤, 소각은 9,524 천톤으로 꾸준히 증가하고 있다. 환경부는 이러한 매립, 소각되는 폐기물 중 약 56%가 재활용 가능한 것으로 발표하였다. 이러한 국내적 상황을 염두 하였을 때, 자원으로 사용가능한 폐기물을 단순 매립 및 소각하는 것은 바람직하지 않다. 많은 선진국은 이미 이러한 폐기물 문제, 자원위기, 에너지 및 환경에 대한 문제를 극복하기 위해 1990년대부터 자원순환사회로 전환하고 있다. 특히, 오스트리아, 프랑스, 폴란드, 스웨덴, 영국 등 선진국들은 매립세 및 소각세를 도입하여 자원순환사회로의 전환을 꾀하였다. 오스트리아의 경우 오염된 부지를 정화하기 위하여 1989년에 매립세를 도입하였으며, 폐기물 종류에 따라 매립세는 다르게 책정되어 있다. 반응성 폐기물에 대한 매립세는 2003년 43.6 euro/ton에서 2004년 65 euro/ton로 증가하였다. 그 결과 매립률은 30.1%에서 11.8%로 크게 감소하였다. 이후 반응성 폐기물에 대한 매립세는 2005년 65 euro/ton에서 2006년 87 euro/ton으로 증가하였으나 매립률은 11.3%에서 9.9%로 약간 감소하였다. 따라서 매립률 감소를 위한 매립세는 적정하게 책정되어야 한다. 환경부에서는 폐기물의의 발생 억제 및 발생된 폐기물의 순환이용과 적정한 처분을 촉진하여 환경을 보전하고 지속가능한 자원순환사회를 만들기 위해 자원순환기본법이 제정하였으며, 2018년 1월 1일부터 시행될 예정이다. 자원순환기본법의 주요 내용으로 자원순환 기반 구축, 자원순환 촉진 수단, 자원순환사업지원 등이 있다. 이중 자원순환 촉진 수단으로는 재활용외의 매립 및 소각 폐기물에 부담금을 책정함으로써 재활용 비용보다 매립 및 소각비용을 더 비싸게 하여 매립 및 소각을 억제하는 방안이 포함되어 있다. 환경부에서는 폐기물 종류에 따라 매립 시 10~30원/kg, 소각 시 10원/kg의 폐기물처분부담금(안)을 공표하였다. 본 연구에서는 자원순환기본법에서의 폐기물처분에 대한 적정 부담금 산정 방안을 위하여 국내･외에서 실행하고 있는 폐기물처분 부담금제를 비교하고자 하며, 폐기물처분 부담금에 의한 폐기물 소각･매립･재활용 추세 등을 검토하고자 한다.