Activated carbon (AC) is used for filtering organic and radioactive particles, in liquid and ventilation systems, respectively. Spent ACs (SACs) are stored till decaying to clearance level before disposal, but some SACs are found to contain C-14, a radioactive isotopes 5,730 years halflife, at a concentration greater than clearance level concentration, 1 Bq/g. However, without waste acceptance criteria (WAC) regarding SACs, SACs are not delivered for disposal at current situation. Therefore, this paper aims to perform a preliminary disposal safety examination to provide fundamental data to establish WAC regarding SACs SACs are inorganic ash composed mostly of carbon (~88%) with few other elements (S, H, O, etc.). Some of these SACs produced from NPPs are found to contain C-14 at concentration up to very-low level waste (VLLW) criteria, and few up to low-level waste (LLW) criteria. As SACs are in form of bead or pellets, dispersion may become a concern, thus requiring conditioning to be indispersible, and considering VLL soils can be disposed by packaging into soft-bags, VLL SACs can also be disposed in the same way, provided SACs are dried to meet free water requirement. But, further analysis is required to evaluate radioactive inventory before disposal. Disposability of SACs is examined based on domestic WAC’s requirement on physical and chemical characteristics. Firstly, particulate regulation would be satisfied, as commonly used ACs in filters are in size greater than 0.3 mm, which is greater than regulated particle size of 0.2 mm and below. Secondly, chelating content regulation would be satisfied, as SACs do not contain chelating chemicals. Also, cellulose, which is known to produce chelating agent (ISA), would be degraded and removed as ACs are produced by pyrolysis at 1,000°C, while thermal degradation of cellulose occurs around 350~600°C. Thirdly, ignitability regulation would be satisfied because as per 40 CFR 261.21, ignitable material is defined with ignition point below 60°C, but SACs has ignition point above 350°C. Lastly, gas generation regulation would be satisfied, as SACs being inorganic, they would be targeted for biological degradation, which is one of the main mechanism of gas generation. Therefore, SACs would be suitable to be disposed at domestic repositories, provided they are securely packaged. Further analysis would be required before disposal to determine detailed radioactive inventories and chemical contents, which also would be used to produce fundamental data to establish WAC.

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.

Low- and intermediate level waste (LILW) repository in Gyeongju, Korea is in operation and the radioactive waste should satisfy the waste acceptance criteria (WAC) of the repository. Among the WAC of the Gyeongju LILW repository, the leachability index criterion is considered to be the criterion that is directly related to the isolation of the radionuclides from biosphere. Cesium, strontium, and cobalt should satisfy the leachability index larger than six by following the ANS 16.1 leaching test method. Several research were performed for the leachability index of Cs, Sr, and Co by following the ANS 16.1 leaching test method. However, the test condition of the previous research is expected to be different to the condition of the actual waste. Due to the radioactivity of the radionuclide such as Cs, Sr and Co, most of the research applied the surrogate of those radionuclides. The concentration of those nuclides was generally measured by the inductively coupled plasma (ICP) equipment, however, high concentration compared to the disposal limit of those nuclides due to the detection limit of the ICP was applied. From the Freundlich and Langmuir adsorption isotherms, the adsorption of the nuclides differs according to the concentration of the nuclides. As the leachability index of the nuclides is affected by the adsorption of the nuclides on the binding material, the effect of nuclide concentration is expected to be not ignorable. Therefore, the leachability index difference according to the nuclide concentration should be compared to avoid over- or underestimation of the leachability index. In this study, the difference in the leachability index according to the concentration of nuclides is aimed to be checked. Cs, Sr, and Co, which should satisfy the leachability index criterion in the WAC of the Gyeongju repository, were selected as target nuclides. Three concentrations were selected to compare the leachability index: 0.1 mol, 0.001 mol and below the regulatory exemption concentration. The concentration of non-radioactive nuclides in the leachant was measured by ICPOES and ICP-MS while the concentration of radionuclides was measured by HPGe. The result of this study can be applied as background data enhancing the WAC or disposal concentration limit of the radionuclides in Gyeongju LILW repository.

Among the nuclear power plant facility improvement projects, out of a total 10 replacement reactor vessel closure head (RRVCH), five have been replaced, starting with Gori Unit 1, and five, including Hanul Unit 1, Hanbit Units 5 and 6, and Hanul Units 3 and 4 will be replaced in the future. This paper presents the method of treating Latch Housing among radioactive waste generated during the replacement of Hanul Unit 2 (February 2023). Latch Housing controls the control rod by receiving magnetic force from the CRDM’s Coil Stack. Located in the Old Reactor Vessel Head (ORVH) Hot Spot, the range of measured radiation dose rate was 0.3 to 0.8 mSv h-1 (up to 4.5 mSv h-1). The amount of radioactive waste generated was 35.8 Baler-Drum (based on 200L), and the order of treatment was to cut into the Omega Seal of the CRDM, the CRDM and Latch Housing Transfer, the boundary of the CRDM and Latch Housing, the Rod Travel Housing, the Motor Housing and the Latch Assembly, and then transfer and Drumming. In the United States, out of 93 operating reactors, 31 reactor vessel heads have been replaced and 19 reactor vessel heads are scheduled to be replaced. In Korea, 25 reactors are in operation, and two reactors have been permanently shut down. Among them, the nine old reactors for more than 30 years (as of September 2021) are expected to achieve ALARA and reduce radwaste management costs through the management method applied to replace the reactor vessel head.

There is a large amount of radioactive waste in waste storage in the Korea Atomic Energy Research Institute. Some of the radioactive waste was generated during the dismantling process due to Korea Research Reactor 1&2 and it accounts for 20% of the total waste. Radioactive waste must be reduced by appropriate disposal methods to secure storage space and to reduce disposal costs. Research Reactor wastes include wastes that are below the acceptable criteria for selfdisposal and non-contaminated wastes, so they can be treated as wastes subject to self-disposal through contamination analysis and reclassification. In order to deregulation radioactive waste, it is necessary to meet the self-disposal standards stipulated in the Domestic Nuclear Act and the treatment standards of the Waste Management Act. The main factors of deregulation are surface contaminant, radionuclide activity and dose assessment. To confirm the contamination of waste, surface contaminant and gamma nuclide analysis were performed. After homogenizing the waste sample, it was placed in 1 L Mariinelli beaker. When collecting waste samples, 1 kg per 200 kg of waste was collected. The concentrations of the major radionuclides Co-60, Cs-134, Cs-137, Eu-152, and Eu-154 were analyzed using HPGe detector. To evaluate radiation dose, various computational programs were used. A dose assessment was performed with the analyzed nuclide concentration. The concentrations of representative nuclides satisfied the deregulation acceptance criteria and the results of the dose assessment corresponding to self-disposal method was also satisfied. Based on this results, KAERI submitted the report on waste self-disposal plan to obtain approval. After final approval, Research Reactor waste is to be incinerated and incineration ash is to be buried in the designated place. Some metallic waste has been recycled. In this study, the suitability of deregulation for self-disposal was confirmed through the evaluation of the surface contaminant analysis, radionuclide concentration analysis and dose assessment.

The development of existing radioactive waste (RI waste) management technologies has been limited to processing techniques for volume reduction. However, this approach has limitations as it does not address issues that compromise the safety of RI waste management, such as the leakage of radioactive liquid, radiation exposure, fire hazards, and off-gas generation. RI waste comes in various forms of radioactive contamination levels, and the sources of waste generation are not fixed, making it challenging to apply conventional decommissioning and disposal techniques from nuclear power plants. This necessitates the development of new disposal facilities suitable for domestic use. Various methods have been considered for the solidification of RI waste, including cement solidification, paraffin solidification, and polymer solidification. Among these, the polymer solidification method is currently regarded as the most suitable material for RI waste immobilization, aiming to overcome the limitations of cement and paraffin solidification methods. Therefore, in this study, a conceptual design for a solidification system using polymer solidification was developed. Taking into account industrial applicability and process costs, a solidification system using epoxy resin was designed. The developed solidification system consists of a pre-treatment system (fine crush), solidification system, cladding system, and packing system. Each process is automated to enhance safety by minimizing user exposure to radioactive waste. The cladding system was designed to minimize defects in the solidified material. Based on the proposed conceptual design in this paper, we plan to proceed with the specific design phase and manufacture performance testing equipment based on the basic design.

Radioactive iodine-129, a byproduct of nuclear fission in nuclear power plants, presents significant environmental and health risks due to its high solubility in water and volatility. Iodine-129, with its half-life of 1.57×1017 years, necessitates safe management and disposal. Therefore, safely capturing and managing I-129 during spent nuclear fuel reprocessing is of paramount importance. To address these challenges, various glass waste forms containing silver iodide have been developed, such as borosilicate, silver phosphate, silver vanadate, and silver tellurite glasses. These glasses effectively immobilize iodine, but the high cost of silver raises affordability concerns. This study introduces CuI·Cu2O·TeO2 glass waste forms for iodine immobilization, a novel approach. The cost-effectiveness of copper, in contrast to silver, makes it an attractive alternative. The CuI·Cu2O·TeO2 glass waste forms were synthesized with varying CuI content (x) in (1-x)(0.3Cu2O·0.7TeO2) glass matrices. Xray diffraction (XRD) confirmed amorphous structures, and X-ray fluorescence (XRF) quantified composition. X-ray photoelectron spectroscopy (XPS) and Raman spectroscopy provided insights into structural properties. Durability assessments using a 7-day product consistency test (PCT-A) and inductively coupled plasma-mass spectrometry (ICP-MS) revealed compliance with U.S. glass regulations, making CuI·Cu2O·TeO2 glasses a promising choice for iodine immobilization in radioactive waste.

We conducted safety assessments for the disposal of spent resin mixed waste after the removal of beta radionuclides (3H, 14C) in a landfill facility. The spent resin tank of Wolsong nuclear power plant is generated by 8:1:1 weight ratio of spent ion exchange resin, spent activated carbon and zeolite. Waste in the spent resin tank was classified as intermediate-level radioactive waste due to 14C. Other nuclides such as 60Co and 137Cs exhibit below the low-level radioactive waste criteria. The techniques for separating mixed waste and capturing 14C have been under development, with a particular focus on microwave-based methods to remove beta radionuclides (3H, 14C) from spent activated carbon and spent resin within the mixed waste. The spent resin and activated carbon within the waste mixture exhibits microwave reactivity, heated when exposed to microwaves. This technology serves as a means to remove beta isotopes within the spent resin, particularly by eliminating 14C, allowing it to meet the low-level radioactive waste criteria. Using this method, the waste mixture can meet disposal requirements through free water and 3H removal. These assessments considered the human intrusion scenarios and were carried out using the RESRAD-ONSITE code. The institutional management period after facility closure is set at 300 years, during which accidental exposures resulting from human intrusion into the disposal site are accounted for. The assessment of radiation exposure to intruders in a landfill facility included six human intrusion scenarios, such as the drilling scenario, road construction scenario, post-drilling scenario, and post-construction scenario. Among the six human intrusion scenarios considered, the most conservative assessment about annual radiation exposure was the post-drilling scenario. In this scenario, human intrusion occurs, followed by drilling and residence on the site after the institutional management period. We assumed that some of the vegetables and fruits grown in the area may originate from contaminated regions. Importantly, we confirmed that radiation doses resulting from post-institutional management period human intrusion scenarios remain below 0.1 mSv/y, thus complying with the annual dose limits for the public. This research underscores the importance of effectively managing and securing radioactive waste, with a specific focus on the safety of beta radionuclide-removed waste during long-term disposal, even in the face of potential human intrusion scenarios beyond the institutional management period.

Structural stability of a waste form can be provided by the waste form itself (steel components, etc.), by processing the waste to a stable form (solidification, etc.), or by emplacing the waste in a container or structure that provides stability (HICs or engineered structure, etc.). The waste or container should be resistant to degradation caused by radiation effects. In accordance with the requirements for the domestic waste acceptance criteria, irradiation testing of solidified waste forms containing spent resin should be conducted on specimens exposed to a dose of 1.0E+6 Gy and other material 1.0E+7 Gy. Expected cumulative dose over 300 years is about 1.770E+6 Gy for spent resin and 0.770E+6 Gy for dried concentrated waste generated from NPPs generally. According to NRC Waste Form Technical Position, to ensure that spent resins will not undergo adverse degradation effects from radiation, resins should not be generated having loadings that will produce greater than 1E+6 Gy total accumulated dose. If it necessary to load resins higher than 1E+6 Gy, it should be demonstrated that the resin will not undergo radiation degradation at the proposed higher loading. This is the recommended maximum activity level for organic resins based on evidence that while a measurable amount of damage to the resin will occur at 1E+6 Gy, the amount of damage will have negligible effect on disposal site safety. Cementitious materials are not affected by gamma radiation to in excess of 1E+6 Gy. Therefore, for cement-stabilized waste forms, irradiation qualification testing need not be conducted unless the waste forms contain spent resins or other organic media or the expected cumulative dose on waste forms containing other materials is greater than 1E+7 Gy. Testing should be performed on specimens exposed to IE+6 Gy or the expected maximum dose greater than 1E+6 Gy for waste forms that contain ion exchange resins or other organic media or the expected maximum dose greater than 1E+7 Gy for other waste forms. This is suggestion as a review result that requirement for irradiation testing of solidified waste forms has something to be revise in detail and definitively.

Among nuclear power plants in the Republic of Korea, Kori Unit 1 and Wolsong Unit 1 have been permanently shut down, and Kori Unit 1 is preparing to be decommissioned. According to the decommissioning plan (DP) of Kori Unit 1, a radioactive waste processing complex will be built on the Kori site to reduce radioactive waste generated during decommissioning actively, and various types of decommissioning waste are expected to be treated in the complex. It is judged that matters related to the safety assessment of the complex are not included in the DP since the equipment and treatment processes have not been determined. IAEA GSR Part 5 states that radioactive waste processing complex shall be operated according to national regulations and the conditions imposed by the regulatory body. However, it has been confirmed that separate regulatory requirements for the complex have not yet been established in Korea. It is expected that the Regulation on Technical Standards for Nuclear Facilities, etc. will be applied mutatis mutandis. Liquid and gaseous radioactive materials can be expected to be released into the sea or atmosphere during the operation of the complex. Accordingly, it should be proved that standards such as discharge limits of radioactive effluents are met. Although the assessment of radioactive effluent discharged from nuclear power plants to the environment is systematically conducted, it has been confirmed that the safety assessment framework for radioactive effluents discharged from the complex has not yet been established. Currently, the SAFRAN Tool is based on SADRWMS (Safety Assessment Driving Radioactive Waste Management Solutions), an IAEA safety assessment methodology for pre-disposal management, which uses Pathway Dose Factors (PDFs) derived from generic environmental models. Therefore, in order to conduct a more detailed safety assessment of the complex on a specific site, site characteristic data should be reflected. Although safety assessment using the SAFRAN Tool was conducted at the Thailand Institute of Nuclear Technology (TINT) facility, detailed data were not provided, and PDFs reflecting site characteristic data were not applied. Also, no other studies that considered many types of waste and provided detailed data on the safety assessment were not confirmed. Therefore, this study developed K-CRAFT (Kyung Hee – Comprehensive RAdioactive waste treatment Facility safety assessment Tool), this tool that can derive PDFs by reflecting site characteristic data based on the SADRWMS methodology and conducted preliminary safety assessment for the complex which will be built in Kori site by this tool.

Properties of bentonite, mainly used as buffer and/or backfill materials, will evolve with time due to thermo-hydro-mechanical-chemical (THMC) processes, which could deteriorate the long-term integrity of the engineered barrier system. In particular, degradation of the backfill in the evolution processes makes it impossible to sufficiently perform the safety functions assigned to prevent groundwater infiltration and retard radionuclide transport. To phenomenologically understand the performance degradation to be caused by evolution, it is essential to conduct the demonstration test for backfill material under the deep geological disposal environment. Accordingly, in this paper, we suggest types of tests and items to be measured for identifying the performance evolution of backfill for the Deep Geological Repository (DGR) in Korea, based on the review results on the performance assessment methodology conducted for the operating license application in Finland. Some of insights derived from reviewing the Finnish case are as follows: 1) The THMC evolution characteristics of backfill material are mainly originated from hydro-mechanical and/or hydrochemical processes driven by the groundwater behavior. 2) These evolutions could occur immediately upon installation of backfill materials and vary depending on characteristics of backfill and groundwater. 3) Through the demonstration experiments with various scales, the hydro-mechanical evolution (e.g. advection and mechanical erosion) of the backfill due to changes in hydraulic behavior could be identified. 4) The hydro-chemical evolution (e.g. alteration and microbial activity) could be identified by analyzing the fully-saturated backfill after completing the experiment. Given the findings, it is judged that the following studies should be first conducted for the candidate backfill materials of the domestic DGR. a) Lab-scale experiment: Measurement for dry density and swelling pressure due to saturation of various backfill materials, time required to reach full saturation, and change in hydraulic conductivity with injection pressure. b) Pilot-scale experiment: Measurement for the mass loss due to erosion; Investigation on the fracture (piping channel) forming and resealing in the saturation process; Identification of the hydro-mechanical evolution with the test scale. c) Post-experiment dismantling analysis for saturated backfill: Measurement of dry density, and contents of organic and harmful substances; Investigation of water content distribution and homogenization of density differences; Identification of the hydro-chemical evolution with groundwater conditions. The results of this study could be directly used to establishing the experimental plan for verifying performance of backfill materials of DGR in Korea, provided that the domestic data such as facility design and site characteristics (including information on groundwater) are acquired.

The effect of various physicochemical processes, such as seawater intrusion, on the performance of the engineered barrier should be closely analyzed to precisely assess the safety of high-level radioactive waste repository. In order to evaluate the impact of such processes on the performance of the engineered barrier, a thermal-hydrological-chemical model was developed by using COMSOL Multiphysics and PHREEQC. The coupling of two software was achieved through the application of a sequential non-iterative approach. Model verification was executed through a comparative analysis between the outcomes derived from the developed model and those obtained in prior investigations. Two data were in a good agreement, demonstrating the model is capable of simulating aqueous speciation, adsorption, precipitation, and dissolution. Using the developed model, the geochemical evolution of bentonite buffer under a general condition was simulated as a base case. The model domain consists of 0.5 m of bentonite and 49.5 m of granite. The uraninite (UO2) was assigned at the canister-bentonite interface as the potential source of uranium. Assuming the lifetime of canister as 1,000 years, the porewater mixing without uranium leakage was simulated for 1,000 years. After then, the uranium leakage through the dissolution of uraninite was initiated and simulated for additional 1,000 years. In the base case model, where the porewater mixing between the bentonite and granite was the only considered process, the gypsum tended to dissolve throughout the bentonite, while it precipitated in the vicinity of bentonite-granite boundary. However, the precipitation and dissolution of gypsum only showed a limited effect on the performance of the bentonite. Due to the low solubility of uraninite in the reduced environment, only infinitesimal amounts of uranium dissolved and transported through the bentonite. Additional cases considering various environmental processes, such as seawater or cement porewater intrusion, will be further investigated.

The operation time of a disposal repository is generally more than one hundred years except for the institutional control phase. The structural integrity of a repository can be regarded as one of the most important research issues from the perspective of a long-term performance assessment, which is closely related to the public acceptance with regard to the nuclear safety. The objective of this study is to suggest the methodology for quantitative evaluation of structural integrity in a nuclear waste repository based on the adaptive artificial intelligence (AI), fractal theory, and acoustic emission (AE) monitoring. Here, adaptive AI means that the advanced AI model trained additionally based on the expert’s decision, engineering & field scale tests, numerical studies etc. in addition to the lab. test. In the process of a methodology development, AE source location, wave attenuation, the maximum AE energy and crack type classification were subsequently studied from the various lab. tests and Mazars damage model. The developed methodology for structural integrity was also applied to engineering scale concrete block (1.3 m × 1.3 m × 1.3 m) by artificial crack generation using a plate jacking method (up to 30 MPa) in KURT (KAERI Underground Research Tunnel). The concrete recipe used in engineering scale test was same as that of Gyeongju low & intermediate level waste repository. From this study, the reliability for AE crack source location, crack type classification, and damage assessment increased and all the processes for the technology development were verified from the Korea Testing Laboratory (KTL) in 2022.

The increasing accumulation of spent nuclear fuel has raised interest in High-Level Waste (HLW) repositories. For example, Sweden is under construction of the KBS-3 repository. To ensure the safety of such HLW repository, various countries have been developing assessment models. In the Republic of Korea, the Korea Atomic Energy Research Institute has been developing on the AKRS model. However, traditional safety assessment models have not considered the fracture growth in the far-field host rock as a function of time. As repository safety assessments guarantee safety for million years, sustained stress naturally leads to the progressive growth of fractures as time goes on. Therefore, it becomes essential to account for fracture growth in the surrounding host rock. To address this, our study proposes a new coupling scheme between the Fracture growth model and the radionuclide transport model. That coupling scheme consists of the Cubic Law model as a fracture growth function and the GoldSim code which is a commercial software for radionuclide transport calculations. The model that adopting such fracture growth functions showed an increase of up to 15% in the release of radionuclide compared to traditional assessment models. our observations indicated that crack growth as a function of time led to an increase in hydraulic conductivity that allowed more radionuclide transport. Notably, these findings show the significance of adopting fracture growth models as a critical element in evaluating the safety of nuclear waste repositories.

Understanding the long-term geochemical evolution of engineered barrier system is crucial for conducting safety assessment in high-level radioactive waste disposal repository. One critical scenario to consider is the intrusion of seawater into the engineered barrier system, which may occur due to global sea level rise. Seawater is characterized by its high ionic strength and abundant dissolved cations, including Na, K, and Mg. When seawater infiltrates an engineered barrier, such dissolved cations displace interlayer cations within the montmorillonite and affect to precipitation/ dissolution of accessory minerals in bentonite buffer. These geochemical reactions change the porewater chemistry of bentonite buffer and influence the reactive transport of radionuclides when it leaked from the canister. In this study, the adaptive process-based total system performance assessment framework (APro), developed by the Korea Atomic Energy Research Institute, was utilized to simulate the geochemical evolution of engineered barrier system resulting from seawater intrusion. Here, the APro simulated the geochemical evolution in bentonite porewater and mineral composition by considering various geochemical reactions such as mineral precipitation/dissolution, temperature, redox processes, cation exchange, and surface complexation mechanisms. The simulation results showed that the seawater intrusion led to the dissolution of gypsum and partial precipitation of calcite, dolomite, and siderite within the engineered barrier system. Additionally, the composition of interlayer cation in montmorillonite was changed, with an increase in Na, K, and Mg and a decrease in Ca, because the concentrations of Na, K, and Mg in seawater were 2-10 times higher than those in the initial bentonite porewater. Further studies will evaluate the geochemical sorption and transport of leaked uranium-238 and iodine-129 by applying TDB-based sorption model.

With the importance of permanent disposal of high-level radioactive waste (HLW) generated in Korea, the deep geological disposal system based on the KBS-3 type is being developed. Since the deep geological repository must provide the long-term isolation of HLW from the surface environment and normal habitats for humans, plants, and animals, it is essential to assess the longterm performance of the disposal facility considering thermal-hydraulic-mechanical-chemical (TH- M-C) evolution. Decay heat dissipated from HLW contained in the canister causes an increase in temperature in the adjacent area. The requirement for the maximum temperature is established in consideration of the possibility of bentonite degradation. Therefore, when designing the repository, the temperature in the region of interest should be identified in detail through the thermal evolution assessment to ensure that the design requirement is satisfied. In the thermal evolution analysis, it is needed to evaluate the temperature distribution over the entire area of the disposal panel to consider the heat generated from both a single canister and adjacent canisters. Computational fluid dynamics (CFD) codes are widely used for detailed temperature analysis but are limited to simulating a wide range. Accordingly, in this study, we developed an analytical solution-based program for efficiently calculating the temperature distribution throughout the deposition panel, which is based on threedimensional heat conduction equations. The code developed can assess the temperature distribution of engineered and natural barrier systems. Principal parameters to be inputted are as follows: (a) geometry of the panel (e.g. width, length, height, spacing between canisters), (b) geometry of the canister (e.g. diameter, height), (c) thermal properties of bentonite and host-rock, (d) initial conditions (e.g. residual heat, temperature), and (e) time information (e.g. canister emplacement rate, time-interval, period). Through the calculation for the conceptual problem of a deposition panel capable of accommodating 900 (i.e. 30×30) canisters, it was confirmed that the program can adequately predict when and where the maximum temperature will occur. It is expected that the overall temperature distribution within the panel can be obtained by the evaluation of the entire region using this program reflecting the detailed design of the repository to be developed in the future. In addition, the thermal evolution analysis considering the influence of other canisters can be performed by applying the results as boundary conditions in the CFD analysis.

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.

In Natural Analogue Study, Concrete is one of the important engineering barrier components in the Multi-thin wall concept of radioactive waste disposal and plays the most important role in disposal sites. The concrete barrier at the disposal site loses its role as a barrier due to various deterioration phenomena such as settlement, earthquake, and ground movement, causing the disposed waste to leak into the natural ecosystem. Some of the key factor is deterioration triggered by sulfate attack. Concrete deterioration induced by sulfate is commonly manifested in an extensive scale when a concrete structure makes contact with soil or water, aggravating its performance. In this study, an accelerated concrete deterioration evaluation experiment was performed using a total of three experimental methods to evaluate the reaction between concrete and water. The first experiment was a deterioration evaluation using Demi. Water, the second was a deterioration evaluation using KURT groundwater after extraction, and the last experiment was a concrete deterioration evaluation using KURT groundwater and sodium sulfate. For all of these experiments, accelerated concrete deterioration experiments were conducted after immersion for a total of 365 days, and specimens were taken out at 30-day intervals and characterization analysis such as SEM and EDS was performed. Experimental analyzes have shown that various chemical species are generated or destroyed over time. In the future, we plan to continue to conduct a total of three concrete deterioration evaluation experiments above, and additionally evaluate the chemical reaction between bentonite and concrete.

The design and fabrication of suitable waste forms with high thermal and structural stability are essential for the safe management and disposal of radioactive wastes. In particular, the thermal properties and temperature distribution of waste form containing high heat-generating nuclides such as Cs and Sr can be used to evaluate its thermal stability, but also provide useful information for the design of canisters, storage systems, and repositories. In this study, a new program code-based thermal analysis framework has been developed to facilitate the characterization, design, and optimization of the waste form. Matlab was used as a software development platform because it provides powerful mathematical computation and visualization components such as the partial differential equation (PDE) toolbox for solving heat transfer problems using finite element method, the App Designer for graphical user interface (GUI), and the MATLAB Compiler for sharing MATLAB programs as standalone applications and web applications. The thermal analysis results such as temperature distribution, heat flux, maximum/ minimum temperature, and centerline/surface temperature profile are visualized with graphs and tables. To evaluate the effectiveness of the developed program, several design and optimization studies were carried out for the SrTiO3 waste form, selected as a stable form of strontium nuclide.

Currently, the most promising fuel candidate for use in sodium fast reactors (SFRs) is metallic fuel, which is produced by a modified casting method in which the metallic fuel material is sequentially melted in an inert atmosphere to prevent volatilization, followed by melting in a graphite crucible, and then injection casting in a quartz (SiO2) mold to produce metallic fuel slugs. In previous studies, U-Zr metallic fuel slugs have been cast using Y2O3 reaction prevent coatings. However, U-Zr alloy-based metallic fuel slugs containing highly reactive rare earth (RE) elements are highly reactive with Y2O3-coated quartz (SiO2) molds and form a significant thickness of surface reaction layer on the surface of the metallic fuel slug. Cast parts that have reacted with nuclear fuel materials become radioactive waste. To decrease amount of radioactive waste, advanced reaction prevent material was developed. Each RE (Nd, Ce, Ln, Pr) element was placed on the reaction prevent material and thermal cycling experiments were carried out. In casting experiments with U-10wt% Zr, it was reported that Y2O3 layer has a high reaction prevent performance. Therefore, the reaction layer properties for RE elements with higher reactivity than uranium elements were evaluated. To investigate the reaction layer between RE and NdYO3, the reaction composition and phase properties as a function of RE content and location were investigated using SEM, EDS, and XRD. The results showed that NdYO3 ceramics had better antireaction performance than Y2O3.