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.

Spent nuclear fuel temporary storage in South Korea is approximately 70% of total storage capacity as of the 4th quarter of 2022 amount is stored. In addition, according to the analysis of the Korean Radioactive Waste Society, saturation of nuclear power plant temporary storage is expected sequentially from 2031, and accordingly, the need for high-level radioactive waste disposal facilities has emerged. Globally, after the conclusion of the EU Taxonomy, for nuclear energy in order to become an ecofriendly energy, it is necessary to have a high-level radioactive waste disposal site and submit a detailed operation plan for high-level radioactive waste disposal site by 2050. Finland and Sweden have already received permission for the construction of high-level radioactive waste disposal facilities, and other countries, such as Switzerland, Japan, the United States, and Canada, are in the process of licensing disposal facilities. In order to establish a repository for high-level radioactive waste, the performance and safety analysis of the repository must be conducted in compliance with regulatory requirements. For safety analysis, it needs a collection of arguments and evidence. and IAEA defined it as ‘Safety case’. The Systematic method, which derives scenarios by systematizing and combining possible phenomena around the repository, is widely used for developing Safety case. Systematic methods make use of the concept of Features, Events and Processes (FEP). FEP identifies features that affect repository performance, events that can affect a short period of time, and processes that can have an impact over a long period of time. Since it is a characteristic of the Systematic method to compose a scenario by combining these FEP, the Systematic method is the basic premise for the development of FEP. Completeness is important for FEP, and comprehensiveness is important for scenarios. However, combining all the FEP into one scenario is time-consuming and difficult to ascertain the comprehensiveness of the scenario. Therefore, an Integrated FEP list is being developed to facilitate tracking between FEP and scenarios by integrating similar FEP. In this study, during the integrated FEP development process, a method for utilizing experts that can be used for difficult parts of quantitative evaluation and a quantitative evaluation process through the method were presented.

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.

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.

As Korea has relatively small land area and large population density compared to other countries considering the DGD concept such as Finland and Sweden, improvements of disposal efficiency in the viewpoint of the disposal area might be needed for the current disposal system to alleviate the difficulties of site selection for the HLW repository. In this research, we conduct a numerical investigation of the disposal efficiency enhancement for a high-level radioactive waste (HLW) repository through three design factors: decay heat optimization, increased thermal limit of buffer, and double-layer concept. In the optimized decay heat model, seven SNFs with the maximum and minimum decay heat depending on actual burn-up and cooling time are iteratively combined in a canister. Thermal limit of buffer is assumed as 100°C and 130°C for reference and high-efficiency repository concepts, respectively. By implementing an optimized decay heat model and a single-layer concept with a thermal limit of buffer set at 100°C, the disposal efficiency increases to 2.3 times of the improved Korean Reference disposal System (KRS+). Additionally, incorporating either an increased thermal limit of buffer to 130°C or a double-layer concept leads to a further 50% improvement in disposal efficiency. By integrating all three design factors, the disposal efficiency can be enhanced up to five times that of the KRS+ repository. Our analysis of rock mass stability reveals that increasing the thermal limit of buffer can generate rock spalling failure in a wider area. However, when accounting for the effect of confining stress by swelling of buffer and backfill using the Mohr-Coulomb failure criteria, the rock mass failure only occurred at the corner between the disposal tunnel and deposition hole when the thermal limit of buffer was increased and a single-layer concept was applied. The results given in this study can provide various options for designing the high-efficiency repository in accordance with the target disposal area and quality of the rock mass in the potential repository site.

Several countries have been operating radioactive waste disposal (RWD) programs to construct their own repositories and have used natural analogues (NA) studies directly or indirectly to ensure the reliability of the long-term safety of deep geological disposal (DGD) systems. A DGD system in Korea has been under development, and for this purpose a generic NA study is necessary. The Korea Atomic Energy Research Institute has just launched the first national NA R&D program in Korea to identify the role of NA studies and to support the safety case in the RWD program. In this article, we review some cases of NA studies carried out in advanced countries considering crystalline rocks as candidate host rocks for high-level radioactive waste disposal. We examine the differences among these case studies and their roles in reflecting each country’s disposal repository design. The legal basis and roadmap for NA studies in each country are also described. However because the results of this analysis depend upon different environmental conditions, they can be only used as important data for establishing various research strategies to strengthen the NA study environment for domestic disposal system research in Korea.

Because of the massive development of nuclear power plants in China in recent years, China is facing the challenge of radioactive waste disposal. China has established complete regulatory requirements for radioactive waste disposal, but it also has encountered problems and challenges in low-level radioactive waste disposal in terms of management, selection of disposal facility sites, and implementation of a site selection plan. Three low-level radioactive waste disposal facilities that have been operated in China are described, and their activity limits, locations, and capacities are also outlined. The connotations of “regional” and “centralized” disposal policies are discussed in light of the characteristics of the radioactive waste. The characteristics and advantages of the regional and centralized disposal policies are compared. It is concluded that the regional disposal policy adopted in 1992 can no longer meet the current disposal needs, and China should adopt a combination of the two disposal policies to solve the problem of radioactive waste disposal.

Technology for high-level-waste disposal employing a multibarrier concept using engineered and natural barrier in stable bedrock at 300–1,000 m depth is being commercialized as a safe, long-term isolation method for high-level waste, including spent nuclear fuel. Managing heat generated from waste is important for improving disposal efficiency; thus, research on efficient heat management is required. In this study, thermal management methods to maximize disposal efficiency in terms of the disposal area required were developed. They efficiently use the land in an environment, such as Korea, where the land area is small and the amount of waste is large. The thermal effects of engineered barriers and natural barriers in a high-level waste disposal repository were analyzed. The research status of thermal management for the main bedrocks of the repository, such as crystalline, clay, salt, and other rocks, were reviewed. Based on a characteristics analysis of various heat management approaches, the spent nuclear fuel cooling time, buffer bentonite thermal conductivity, and disposal container size were chosen as efficient heat management methods applicable in Korea. For each method, thermal analyses of the disposal repository were performed. Based on the results, the disposal efficiency was evaluated preliminarily. Necessary future research is suggested.

Concrete waste generated in the result of dismantling a concrete structure in a radiation control area and refractory brick waste generated from uranium pellet sintering furnace are surface-contaminated by uranium particle of which the enrichment is below 5%. These wastes are hard to decontaminate so it was necessary to develop the process for its disposal. So, we developed the Process Control Plan (PCP) for disposal of radioactive concrete waste describing a whole sequence of disposal and inspecting procedures based on the KNF Radioactive Waste Quality Assurance Plan (KN-WQAP) established in 2021. Based on the PCP, we crushed the concrete waste by jaw-crusher. Then we sieved the crushed concrete waste and removed the particle of which size is below 0.3 mm, using sieve-vibrator where the 0.3 mm mesh-sized sieve is installed inside. Before conducting the crush-sieving method based on the PCP, we conducted Process Control Assessment (PCA) based on the KN-WQAP. The purpose of the PCA is to check whether the output of the process satisfies the Acceptance Criteria of Korea Radioactive Waste Agency (KORAD) so that we could confirm the validity of the PCP. The evaluation item of the PCA is a particulate size verification test. The test is passed only if the component ratio of a particle size below 0.2 mm is less than 15% and the particle size below 0.01 mm is less than 1%. The very first 3 drums passed the test, so we began applying the PCP to whole target drums. In the process of conducting the crush-sieving method in earnest, qualified inspectors based on KNWQAP participated conducting sampling, measuring and checking whether a foreign material was included. They tested samples and packaged drums regarding 5 spheres of general, radiological, physical, chemical and biological characteristic. KNF disposed concrete and refractory brick waste by the crush-sieving method so that KNF could take over 100 drums to KORAD in 2021. But, it is needed to be improved that a dust size below 0.3 mm is generated as a secondary waste which needs to be solidified for the final disposal and the work environment is not good enough because of the dust.

The Korean administration assumed that the amount of low and medium level waste generated during the decommissioning of nuclear facilities in Korea was 14,500 drums (based on 200 L) and designed the LILW repository accordingly. Accordingly, it is necessary to separate the nuclear power plant decommissioning waste into clearance waste by mobilizing means such as decontamination and cutting as much as possible, and to deregulate it together with non-radioactive waste. As a result, clearance waste and non-radioactive waste are dominated by concrete and metal, and it is necessary to evaluate how to recycle them. Many existing studies have conducted research on each recycling method, and accordingly, it can be judged that the technological maturity is sufficient. Accordingly, we would like to propose a method for comprehensive management and evaluation of concrete. By applying the decision matrix proposed in IAEA TRS No. 401, it will be possible to compare the 5 factors (cost, technical feasibility, risk, availability of disposal, and full cycle impact). However, in the case of concrete, if the existing construction waste recycling methodologies are fully used, the technical feasibility can be considered equal. Therefore, it was judged that it would be good to introduce the aspect of public acceptance as an evaluation item instead of technical feasibility. The amount of waste that can be generated when decommission a nuclear power plant is only insignificant compared to the total amount of waste concrete that is generated during the year. Accordingly, one option is to fully integrate the waste concrete recycling system and utilize it for road construction. Next, it is possible to suggest the option of recycling in the construction of shields in the nuclear industry, as suggested in previous studies, and the method of using it as a backfill material such as for a decommissioned NPP site or other sites. As an example, and a draft stage, this study was evaluated based on existing studies after all options were equally weighted. When the profit and loss was evaluated in a way that a maximum of 5 points were given to each option, the case of using it as a backfill in various applications was evaluated as the best option. Unlimited recycling, such as road construction, was evaluated to be highly damaging in terms of public acceptance.

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.

This study introduces the licensing process carried out by the regulatory body for construction and operation of the 2nd phase low level radioactive waste disposal facility in Gyeongju. Also, this study presents the experience and lessons learned from this regulatory review for preparing the license review for the next 3rd phase landfill disposal facility. Korea Radioactive Waste Agency (KORAD) submitted a license application to Nuclear Safety and Security commission (NSSC) on December 24, 2015 to obtain permit for construction and operation of the national engineered shallow land disposal facility at Wolsong, Gyeongju. NSSC and Korea Institute of Nuclear Safety (KINS) started the regulatory review process with an initial docket review of the KORAD application including Safety Analysis Report, Radiological Environmental Report and Safety Administration Rules. After reflecting the results of the docket review, the safety review of revised 10 application documents began on November 29, 2016. Total 856 queries and requests for additional information were elicited by thorough technical review until November 16, 2021. As the Gyeongju and Pohang earthquakes occurred in September 2016 and November 2017, respectively, the seismic design of the disposal facility for vault and underground gallery was enhanced from 0.2 g to 0.3 g and the site safety evaluation including groundwater characteristics was re-investigated due to earthquake-induced fault. Also, post-closure safety assessments related to normal/abnormal/human intrusion scenarios were re-performed for reflecting the results of site and design characteristics. Finally, NSSC decided to grant a license of the 2nd phase low level radioactive waste disposal facility under the Nuclear Safety Laws in July 2022. This study introduces important issues and major improvements in terms of safety during the review process and presents the lessons learned from the experience of regulatory review process.

A radioactive waste disposal facility needs to be developed in a way to protect present and future generations and its environment. A safety assessment is implemented for normal and abnormal scenarios and human intrusion scenarios as a part of a safety case in developing a disposal facility for the radioactive waste. The human intrusion scenarios include a well scenario which takes into account various potential exposure groups (PEGs) who use a groundwater well contaminated with radionuclides released from the disposal facility. It is observed that a pumping rate has a negative correlation with the biosphere dose conversion factor (BDCF) in the well scenario. C-14 is shown to be a key radionuclide in the well scenario, and a special model based on the carbon cycle is applied for C-14. For Tc-99, an adsorption coefficient should be adjusted to be suitable for the site. The safety assessment for the radioactive waste disposal facility is successfully carried out for the well scenario. However, it is observed that site-specific models needs to be developed and sitespecific input data need to be collected in order to avoid unnecessary conservatism.

To efficiently manage the waste packages like drums, it is meaningful to utilize the data of placement and emplacement of disposed waste in geological storage. For the transparent and real-time management of disposal data, both technical as well as administrative factors must be included. To this end, MIRAE-EN Co., Ltd. has developed a radioactive waste tracking and management system (m-trekⓇ v1.0) through radioactive waste management life cycle which is supported by KETEP. Enhancing the functional features of m-trekⓇ, IoT-based drum location measuring and data of those drums, such as position, radionuclides, activity, and dose etc., are to be collected and monitored through data modeling and visualization, which might be utilized in emplacing the loaded drums according to specifically certain criteria of internal and external data of disposal site. Position measuring using Beacon utilizes Received Signal Strength Indicator (RSSI) to locate the correct position in 3D area. Since RSSI is affected by the surrounding environment, it is required to corrective optimization. In addition, error and deviation of previously applied Gaussian filter method, was corrected and improved through AI learning model. According to this location information and those data, the prototype in future provides the visualization of drum position and its relevant data for administrative purpose such as monitoring, emplacement and other management policy.

Organic complexing agents may affect the mobility of radionuclides at low- and intermediate-level radioactive waste repositories. Especially, isosaccharinic acid (ISA) is the main cellulose degradation product under high pH conditions in cement pore water. ISA can combine with radionuclides and form stable complexes that adversely influence adsorption in the concrete phase, resulting in radionuclides to leach to the near- and far-fields of repositories. This study focuses on investigating the sorption of ISA onto engineered barriers such as concrete, thereby studying adsorption isotherms of ISA on concrete and comparing various isotherm models with the experimental data. The adsorption experiment was conducted in three background solutions, groundwater (adjusted to pH 13 using NaOH), State 1 (artificial cement pore water, pH 13.3), and State 2 (artificial cement pore water, pH 12.5), in a batch system at a temperature of 20°C. Concrete was characterized using BET, Zeta-potential analyzer, XRD, XRF, and SEM-EDS. ISA concentrations were detected using HPLC. The experimental data were best fitted to one-site Langmuir isotherm; On the other hand, either two-site isotherm or Freundlich isotherm couldn’t give reasonable fitting to the experimental data. The observed ISA sorption behavior on concrete is crucial for the disposal of radioactive waste because it can significantly lower the concentration of ISA in the pore water. Although one-site Langmuir isotherm might effectively represent the sorption behavior of ISA on concrete, the underlying mechanism is still unknown, and further investigation should be done in the near future.

The fuel fabrication facility has been built and is being operated by KAERI since licensing research reactor fuel fabrication in 2004. After almost 20 years of operation, outdated equipment for fabrication or inspection has been replaced by automated, digitalized ones to assure a higher quality of nuclear fuels. However, the generation of a large amount of radioactive waste is another concern for the replacement in terms of its volume and various types of it that should be categorized before disposal. The regulatory body, NSSC (Nuclear Safety and Security Commission) released a notice related to the classification of radioactive wastes, and most accessory equipment can be classified into the clearance levels, called self-disposal waste. In this study, the practice of self-disposal of metal radioactive waste is carried out to reduce its volume and downgrade its radioactivity. For metal radioactive waste, which is expected to occupy the most amount, analysis status and legal limitations were performed as follows: First, the disposal plan was established after an investigation of the use history for equipment. Second, those were classified by types of materials, and their surface radio-contamination was measured for checking self-disposable or not. After collecting data, the plan for the self-disposal was written and submitted to the Korea Institute of Nuclear Safety (KINS) for approval.

Glass fiber (GF) insulation is a non-combustible material, light, easy to transport/store, and has excellent thermal insulation performance, so it has been widely used in the piping of nuclear power plants. However, if the GF insulation is exposed to a high-temperature environment for a long period of time, there is a possibility that it may be crushed even with a small impact due to deterioration phenomenon and take the form of small particles. In fact, GF dust was generated in some of the insulation waste generated during the maintenance process. In the previous study, the disposal safety assessment of GF waste was performed under the abnormal condition of the disposal facility to calculate the radiation exposure dose of the public residing/ residents nearby facilities, and then the disposal safety of GF waste was verified by confirming that the exposure dose was less than the limit. However, the revised guidelines for safety assessment require the addition of exposure dose assessment of workers. Therefore, in this study, accident scenarios at disposal facilities were derived and the exposure dose to the workers during the accident was evaluated. The evaluation was carried out in the following order: (1) selection of accident scenario, (2) calculation of exposure dose, (3) comparison of evaluation results with dose limits, and confirmation of satisfaction. The representative accident scenarios with the highest risk among the facility accident were selected as; (a) the fire in the treatment facility, (b) the fire in the storage facility, and (c) fire after a collision of transport vehicles. The internal and external exposure doses of the worker by radioactive plume were calculated at 10m away from the accident point. In evaluation, the dose conversion factors ICRP-72 and FGR12 were used. As a result of the calculation, the exposure dose to workers was derived as about 0.08 mSv, 0.20 mSv, and 0.10 mSv, due to fire accidents (vehicle collision, storage facilities, treatment facilities). These were 0.2%, 0.4%, and 0.2% of the limit, and the radiation risk to workers was evaluated to be very low. The results of this study will be used as basic data to prove the safety of the disposal of GF waste. The sensitivity analysis will be performed by changing the radiation source and emission rate in the future.

Copper is used for deep geological disposal canisters of spent nuclear fuels, because of excellent corrosion resistance in an oxygen-free environment. However, sulfide formation during the long-term exposure under deep geological disposal condition can be harmful for the integrity of copper canisters. Sulfur around the canisters can diffuse along grain boundaries of copper, causing grain boundary embrittlement by the formation of copper sulfides at the grain boundaries. The development of copper alloys preventing the formation of copper sulfides along grain boundaries is essential for the longterm safety of copper canisters. In this research, the mechanisms of copper sulfide formation at the grain boundary are identified, and possible alloying elements to prevent the copper sulfide formation are searched through the first principle calculations of solute atom-vacancy binding energy and the molecular dynamics calculation of grain boundary segregation energy. The comparison with the experimental literature results on the mitigation of copper embrittlement confirmed that the theoretically identified mechanisms of copper sulfide formation and the selected alloy elements are valid. Thereafter, binary copper alloys were prepared by using a vacuum arc melting furnace. Sulfur was added during casting of the copper alloys to induce the sulfide formation. The cast alloys were cold-rolled into a plate after homogenization heat treatment. The microstructure and mechanical property of each alloy were investigated after recrystallization in a vacuum tube heat treatment furnace. The copper alloys developed in this study are expected to contribute in increasing the long-term safety of deep geological disposal copper canisters by reducing the embrittlement caused by the sulfide formation.

To minimize the short-term thermal load on the repository facility, heat generating nuclides such as Cs-137 and Sr-90 should be separated from the spent nuclear fuel for efficiency of repository facility. In particular, Sr-90 must be separated because it generates high heat during the decay process. Recently, Korea Atomic Energy Research Institute (KEARI) is developing a waste burden minimization technology to reduce the environmental burden caused by the disposal of spent nuclear fuel and maximize the utilization of the disposal facility. The technology includes a nuclide management process that can maximize disposal efficiency by selectively separating and collecting major nuclides such as Cs, Sr, I, TRU/RE, and Tc/Se from spent nuclear fuel. Among the major nuclides, Sr nuclides dissolve in chloride phase during the chlorination process of spent nuclear fuel and recovered in the form of carbonate or oxide via reactive distillation. In this process, Ba nuclides are also recovered along with Sr nuclides due to their chemical similarity. In this study, we prepared group II nuclide ceramic waste form, Ba(x)Sr(1-x)TiO3 (x=0, 0.25, 0.5, 0.75, 1), using the solid-state reaction method by considering the various ratio of Sr/Ba nuclides generated from nuclide management process. The established waste form fabrication process was able to produce a stable waste form regardless of the ratio of Sr/Ba nuclides. To evaluate the stability of group II waste form, physicochemical properties such as leaching and thermal properties were evaluated. Also, the radiological properties of the Ba(x)Sr(1-x)TiO3 waste forms with various Sr/Ba ratios were evaluated, and the estimation of centerline temperature was carried out using the experimental thermal property data. These results provided fundamental data for long-term storage and management of group II nuclides waste form.

Operating and decommissioning nuclear power plants generates radioactive waste. This radioactive waste can be categorized into several different levels, for example, low, intermediate, and high, according to the regulations. Currently, low and intermediate-level waste are stored in conventional 200-liter drums to be disposed. However, in Korea, the disposal of intermediate-level radioactive waste is virtually impossible as there are no available facilities. Furthermore, large-sized intermediate- level radioactive waste, such as reactor internals from decommissioning, need to be segmented into smaller sizes so they can be adequately stored in the conventional drums. This segmentation process requires additional costs and also produces secondary waste. Therefore, this paper suggests repurposing the no-longer-used spent nuclear fuel casks. The casks are larger in size than the conventional drums, thus requiring less segmentation of waste. Furthermore, the safety requirements of the spent nuclear fuel casks are severer than those of the drums. Hence, repurposed spent nuclear fuel casks could better address potential risks such as dropping, submerging, or a fire. In addition, the spent nuclear fuel casks need to be disposed in compliance with the regulations for low level radioactive waste. This cost may be avoided by repurposing the casks.