Decommissioning of a nuclear power plant (NPP) generate large amounts of various types of wastes. In accordance with the Nuclear Safety and Security Commission Notice of Korea (No. 2020- 6), they are classified as High Level Waste (HLW), Intermediate Level Waste (ILW), Low Level Waste (LLW), Very Low Level Waste (VLLW) and Exempt Waste (EW) according to specific activities. More than 90% of the wastes are at exempt level, mostly metal and concrete wastes with low radioactivity, of which the concentrations of nuclides is less than the allowable concentration of self-disposal. The self-disposal or recycling of these wastes is widely used worldwide. More than 10,000 drums, based on 200 L drum, are expected to be produced in the decommissioning process of a unit of nuclear power plant. Due to the limited storage capacity of the intermediate & low level waste disposal facility in Gyeongju, recycling and self-disposal of EW are actively recommended in Korea. A variety of scenarios were proposed for recycling and self-disposal of decommissioning metal/ concrete wastes, and a computational program called REDISA was developed to perform the dose evaluation for each recycling and self-disposal scenario. The REDISA computer program can calculate external and internal exposure doses by simulating the exposure pathways from waste generation, thru transport, processing, manufacture, to the final destination of recycling or self-disposal. In this study, the self-disposal scenario was only considered for the dose evaluation. Many studies have been conducted to evaluate the exposure doses of the radioactive waste disposal sites. However, there have been few researches on dose evaluation for self-disposal landfills. In particular, the dose evaluation is important not only during the operation period, but also for a long period after the facility is closed. To this end, we developed a conceptual model for dose evaluation for post-closure scenarios of the self-disposal landfill of decommissioning metal/concrete wastes with reference to the methodology of IAEA-TECDOC-1380. The model incorporates three exposure pathways, including external exposure from contaminated soil, internal exposure by inhalation, and internal exposure by ingestion of water and food grown in contaminated soil. The duration of the dose evaluation is set to 100,000 years after the closure of landfill facility. Co-60 was selected as dominant nuclide, and dose evaluation was performed based on unit specific activity of 1 Bq/g. Exposure doses shall be verified for their application in accordance with the annual dose limit of 10 Sv/yr for self-disposal. As a result, the post-closure scenario of selfdisposal landfills have shown negligible effects on public health, which means that the exposures doses from transportation and operational processes should be considered more carefully for selfdisposal of decommissioning metal/concrete wastes.

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.

Most of the wastes generated when dismantling nuclear power plant were contaminated with lowlevel radioactive materials, therefore, applying a plasma melting system is a good option to dispose of the complex wastes safely. Melting system with plasma technology was developed to dispose single metal or composite objects. Its purpose is to secure final emissions satisfying final treatment conditions by controlling oxidization/ reduction reaction condition in detail during the melting process. A hollow plasma torch applied at plasma melting system could be operated with various plasmaforming gasses such as N2, Air, Ar, O2, and etc. The melting furnace was designed based on a double sealing structure to prevent risk factors; such as leaks, etc. in the reaction condition. The effect of the external air inflow on the melting conditions was minimized by carefully designing the object input device, torch mounting part, final object discharge part, etc.

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.

During and after the construction of LILW disposal facilities, the decrease of groundwater head potential has been monitored. In addition, an increase of the electrical conductivity (EC) has been observed in several monitoring wells installed along the coastal coastline. Monitoring activity for groundwater head potential and hydrogeochemical properties is important to reduce the uncertainty in the evaluation of groundwater flow characteristics. However, the data observed in the monitoring wells are spatial point data, so there is a limit to the dimension. Several researchers evaluated groundwater head potential changes and seawater intrusion (SWI) potential for disposal sites using groundwater flow modeling. In case of groundwater flow modeling results for SWI, there is a spatial limit in directly comparing the EC observed in the monitoring wells with the modeling results. In a recent study, it was confirmed that the response of the long-range ground penetraiing radar (GPR) system was severely attenuated in the presence of saline groundwater. In order to reduce the spatial constraint of the groundwater monitoring wells for SWI, the characteristics of SWI within the disposal facility site by using the the results of a recent study of the long-range GPR system were investigated and evaluated in this study.

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.

Engineered barriers (concrete and grout) in Low- and Intermediate-Level Waste (L/ILW) disposal facilities tend to degrade by groundwater or rainfall water over a long period of time. During the degradation process, radionuclides stored in the disposal facility might be released into the pore water, which can pass through the natural rock barriers (granite and sedimentary rock) and may reach the near-field and far-field. In this transportation, radionuclide might be sorbed onto the engineered and natural rock barriers. In addition, the organic complexing agent such as ethylenediaminetetraacetic acid (EDTA) and α-isosaccharinic acid (ISA), is also present in pore water, which may affect the sorption and mobility of radionuclide. In this study, the sorption and mobility of 90Sr under different conditions such as two pHs (7 and 13), different initial concentrations of organic complexing agents (from 10-5 M to 10-2 M), and solutions (groundwater, pore water, and rainfall water) were investigated in a batch system. The groundwater was collected at the L/ILW disposal facility located at Gyeongju in South Korea. The pore water and rainfall water were artificially made in the laboratory. The concrete, grout, granite, and sedimentary rock samples were collected from the same study sites from where the groundwater was collected. The rock samples were crushed to 53-150 micrometers and were characterized by XRD, XRF, SEM-EDS, BET, and zeta potential analyzer. 90Sr concentration was determined using liquid scintillation counting. The sorption of 90Sr was described by distribution coefficients (Kd) and sorption reduction factor (SRF). In the case of EDTA, the Kd values of 90Sr remained constant from 10-5 M to 10-3 M and tended to decrease at 10-2 M, while in case of ISA the Kd values decreased steadily as the concentration of ISA was increased from 10-5 M to 10-3 M; However, a sudden reduction in the Kd values were observed above 10-2 M. In comparison to EDTA, ISA gave a higher SRF of 90Sr. Therefore, from the above results, it can be concluded that the presence of ISA has a greater effect on the sorption and mobility of radionuclide in the solutions than EDTA, and the radionuclide may reach near- and far-field of the L/ILW disposal facility.

For the disposal of radioactive waste generated from nuclear power plants, characterization of radioactive waste is essential. For characterization, samples of radioactive waste are directly collected or an indirect method is used through X-ray, etc. Through indirect analysis, which is a non-destructive method, the density, filling height, homogeneity and inter structure of the waste container can be analyzed. Currently, foreign institutions are in the process of developing a technology to perform characterization of radioactive waste through indirect analysis. In particular, research on improving internal image accuracy through image analysis techniques, improving measurement methods and enhancing portability for field application is ongoing. Through the review of such technology development trends, it will be utilized in the development of domestic radioactive waste disposal technolgy.

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

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.

Currently, Hanul NPP packages glass fiber classified as particulate waste in plastic packaging bags and stores them in 200 L drums. KORAD’s Waste Acceptance Criteria (WAC) presents that very low-level soil can be immobilized by loading it in a soft bag and then packaging it in a 200 L or 320 L steel drum. As currently accepted method of packaging with soft bag applies to only very low-level soils among the wastes with a risk of dispersion, it is necessary to develop a non-dispersible treatment suitable for the characteristics of other particulate waste in the future. Therefore, in order for Hanul packaging pack to be approved as an alternative method for immobilization of dispersible substances, it is necessary to verify the suitability of the packaging bag. In this paper, whether the glass fiber packaging bag used in Hanul NPP satisfies the characteristic of the soft bag presented in the WAC and the possibility of being considered as a non-dispersible measure for particulate are examined. The soft bag must meet the following requirements: material and structure, shape, drop test, and immersion test. The results of the review are as follows. First, since the glass fiber is already packaged in the drum, only the role of the inner layer, made of polyethylene, having a watertight function may be required. Second, when packaging a drum, the packaging bag is compressed into a shaped frame having an inner size of a 200 L drum, so it is packaged with little empty space in the drum. Third, as a result of a drop test of a packaging pack containing 20 kg of contents from a height of 1.2 m, it was confirmed that there was no leakage of contents. Fourth, the packaging bag was immersed in a 1-m depth water tank for 30-minutes, and the performance corresponding to the IPX7 was satisfied. As a result of reviewing the soft bag characteristic of Hanul glass fiber packaging bag, it is considered that the bag can be used as one of the non-dispersible measures because it meets almost the characteristics required by the WAC. In addition, the acceptance criteria of overseas disposal sites present various secure packaging methods in place of immobilization as a non-dispersible measure for waste containing particulate matter. It is necessary to reflect these overseas cases in the establishment of non-dispersible measures for domestic waste acceptance in the future.

The Deep Borehole Disposal (DBD) method has various advantages, such as minimizing the use of site area and corrosion of the disposal container and improving long-term structural safety. However, it is necessary to review the problems that may occur in various technologies related to the emplacement and retrieval of the disposal container and the sealing of the borehole. Therefore, the purpose of this study is to evaluate the structural integrity of an emplacement and retrieval device (hereinafter, the disposal container connecting device) of a DBD container. The disposal connecting device was evaluated according to ANSI 14.6 and NUREG-0612 standards. The allowable stress should be less than the yield strength under the load condition of 3g. The length of the disposal container connecting device was about 2,900 mm, the diameter was 406 mm, and the weight was about 1.2 tons. In addition, 10 disposal containers weighing up to 2.2 tons were handled. The disposal container connecting device was made of stainless steel, and the maximum operating temperature was about 300°C. For structural evaluation, ABAQUS finite element analysis program was used. The analysis model was modeled only 1/2 part considering symmetry condition. The analysis model was modeled using 410,431 nodes and 344,119 solid elements. Three times load was applied to the weight of the disposal container. Axisymmetric conditions were applied to the symmetrical surface of the disposal container, and vertical restraints were applied to the upper lifting lugs. A surface-to-surface contact condition was applied to the part where the contact occurred. As a result of the analysis, the greatest stress was generated at the part supported by the clamp at the disposal container connector at 168.9 MPa. In the lugs and pins connecting the guide and the connecting device, a stress of 530.1 MPa was generated by shearing. In the bolts of the disposal container connecting device, a stress of 498MPa was generated and the safety margin was 1.73. A stress of 486.1 MPa was generated in the disposal container connecting device, and the safety margin was the smallest 1.16. As a result of the analysis, all components of the disposal container connecting device showed a safety margin of 1.16 or more at the maximum operating temperature and satisfied the allowable stress.

In order to dispose of spent nuclear fuel (SNF) in deep geological repository, source term evaluation considering its specification, enrichment, burnup, cooling time should be performed. In this study, the measured values of Takahama-3 pressurized water reactor SNF (WH 17×17) samples were analyzed with SCALE 6.1/ORIGEN-S and TRITON code calculation results for validation. Unlike the ORIGENS code, TRITON code calculations differed from two-dimensional neutron flux distribution by using the multi-group cross-section library. Both calculation results from ORIGEN-S and TRITON code showed higher errors in 234U, 239Pu, and 241Pu compared to other actinide nuclides. In the case of axial locations of fuel rods in fuel assembly, fuel rods located at the edge of the fuel assembly presented increased errors due to nuclear reaction cross-section. Overall, the ORIGEN-S predictions informed more accurate agreement with the measured results compared with TRITON results. Especially to 235U, 239Pu, and 240Pu radionuclides, ORIGEN-S errors were denoted more than twice as low as the TRITON results. Comparing the calculation results with experimental results implied that the ORIGENS code was more accurate code than the TRITON code for source term evaluation.

In biosphere assessment modeling for the safety assessment of the Wolsong LILW disposal facility, the multi-compartment modeling in which all radionuclides transport is described quantitatively in terms of transfer factors between various environmental compartments has been implemented. In order to reflect the actual transfer mechanisms of 14C in the environment the specific activity (SA) modeling approach can be applied as an alternative to the previous transfer factors (TF) approach. The assumption of full SA equilibrium throughout the terrestrial environment is completely satisfactory for 14C release to the atmosphere if the 12C is emitted as 14CO2. This is the only form that is readily taken up by plants, so that active carbon is incorporated into the plant via photosynthesis at the same rate as stable carbon. Accordingly, the 14C concentration in Bq/g stable carbon is the same in the plant as it is in the air. And animals take up carbon almost entirely through ingestion and the SA ratio in the plant is maintained in the animal. In this study, a specific activity model for 14C was implemented in a GoldSim biosphere assessment model. From the literature survey for existing specific activity models developed, the IAEA model was selected. The farming scenario utilizing well water was simulated and the resulting ingestion dose conversion factors (DCFs) from the IAEA SA model were compared with those of the TF approach. The parameter value for the concentration of stable carbon in the air (gC/m3) is used as 0.20 gC/m3 considering the Suess effect. The dose coefficient for food ingestion used for dose calculations was taken from ICRP-72 as 5.8E-10 Sv/Bq. It was found that the ingestion DCFs of the SA model showed about 3 times lower than those of the TF model in the farming scenario through irrigation of well water, so it is expected that the SA approach could be applied for a more realistic assessment. Though the comparisons were made on the results from the terrestrial ecosystem only in this study, it would be necessary to investigate the applicability of the SA modeling approach for 14C through extensive comparisons and analysis including an aquatic ecosystem, and through parameters survey suitable to the domestic condition.

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.

Despite the increasing interest in Deep Borehole Disposal (DBD) for its capability of minimizing disposal area, detailed research about DBD operation system design should be conducted before the DBD can be implemented. Recently, DBD operation system applying wireline emplacement (WE) technique is under study due to its high flexibility and capability of minimizing surface equipment. In this study, a conceptual WE system, and operation procdure is introduced. The conceptual WE system consists of 3 main stations, which from the top are hoisting station (HS), canister connection station (CCS) and basement (BS). In HS, WE is controlled and monitored. The WE is controlled using wireline drum winch and sheaves, and load on wireline is measured using a load cell. HS also has a pressure control system (PCS), which monitors internal pressure of the system, and a lubricator, which act as housing for joint device, allowing the joint device to be easily inserted into the borehole. The joint device is used to connect the disposal canister to wireline for emplacement/retrieval. In CCS, a rail transporter brings a transport cask containing disposal canisters, then the transport cask is connected to the hoisting system and a PCS in the BS. The main component located at canister station are a sliding shielding door (SSD), and a slip. The SSD is used to prevent canister from falling into borehole during the connecting operation and prevent radiation from BS to affect the workers. The slip is located beneath the SSD and is used to hold the disposal canister before it is lowered into the borehole. In BS, PCS is installed to prevent overflow and blowout of borehole fluid. The PCS consists of wireline pressure valve, christmas tree and BOP, which all are a type of pressure valve to seal the borehole and release pressure inside the borehole. The WE procedure starts with transporting transport cask to CCS. The transport cask is connected to lubricator, and PCS. Joint device is lowered down to be connected with disposal canisters, then pulled up to check the load on the wireline. After the check-up, SSD is opened, and disposal canister is lowered into the borehole. When desired depth is reached, joint device is disconnected and retrieved for next emplacement. In this study, the conceptual deep borehole disposal system design implementing WE technique is introduced. Based on this study, further detailed design could be derived in future, and feasibility could be tested.