Spent nuclear fuel management is a high-priority issue in South Korea, and addressing it is crucial for the country’s long-term energy sustainability. The KORAD (Korea Radioactive Waste Agency) is leading a comprehensive, long-term project to develop a safe and effective deep geological repository for spent nuclear fuel disposal. Within this framework, we have three primary objectives in this work. First, we conducted statistical analysis to assess the inventory of spent nuclear fuel in South Korea as of 2021. We also projected future generation rates of spent nuclear fuels to identify what we refer to as reference spent nuclear fuels. These reference spent nuclear fuels will be used as the design basis spent fuels for evaluating the safety of the repository. Specifically, we identified four types of design basis reference spent nuclear fuels: high and low burnup from PLUS7 (with a 16×16 array) and ACE7 (with a 17×17 array) assemblies. Second, we analyzed radioactive nuclides’ inventory, activities, and decay heats, extending up to a million years after reactor discharge for these reference spent nuclear fuels. This analysis was performed using SCALE/TRITON to generate the burnup libraries and SCALE/ORIGEN for source term evaluation. Third, to assess the safety resulted from potential radioactive nuclides’ release from the disposal canister in future work, we selected safety-related radionuclides based on the ALI (Annual Limit of Intake) specified in Annex 3 of the 2019-10 notification by the NSSC (Nuclear Safety and Security Commission). Conservative assumptions were made regarding annual water intake by humans, canister design lifetime, and aquifer flow rates. A safety margin of 10-3 of the ALI was applied. We selected 56 radionuclides that exceed the intake limits and have half-lives longer than one year as the safety-related radionuclides. However, it is crucial to note that our selection criteria focused on ALI and half-lives. It did not include other essential factors such as solubility limits, distribution coefficients, and leakage processes. So, some of these nuclides can be removed in a specific analysis area depending on their properties.

The WRK (Waste Repository Korea bentonite) compacted bentonite medium has been considered as the appropriate buffer material in the Korean SNF (Spent nuclear fuel) repository site. In this study, hydraulic properties of the WRK compacted bentonite core (4.5 cm in diameter and 1.0 cm in length) as the buffer material were investigated in laboratory experiments. The porosity and the entry pressure of the water saturated core at different confining pressure conditions were measured. The average velocity of water flow in the WRK compacted bentonite core was calculated from results of the breakthrough curves of the CsI aqueous solution and the hydraulic conductivity of the core was also calculated from the continuous flow core experiments. Because various gases could be generated by continuous SNF fission, container corrosion and biochemical reactions in the repository site, the gas migration property in the WRK compacted bentonite core was also investigated in experiments. The gas permeability and the average of gas (H2) in the core at different water saturation conditions were measured. Laboratory experiments with the WRK Compacted bentonite core were performed under conditions simulating the DGR environment (confining pressure: 1.5- 20.0 MPa, injection pressure: 1.0-5.0 MPa, water saturation: 0-100%). The WRK Compacted bentonite core was saturated at various confining pressure conditions and the porosity ranged from 27.5% to 43.75% (average: 36.75%). The calculated hydraulic conductivity (K) of the core using experimental results was 8.69×10-11 cm/s. The gas permeability of the core when the water saturation 0~58 % was ranged of 19.81~3.43×10-16 m2, representing that the gas migration in the buffer depends directly on the water saturation degree of the buffer medium. The average gas velocity in the core at 58% of water saturation was 9.8×10-6 m/s, suggesting that the gas could migrate fast through the buffer medium in the SNF repository site. Identification of the hydraulic property for the buffer medium, acquired through these experimental measurements is very rare and is considered to have high academic values. Experimental results from this study were used as input parameter values for the numerical modeling to simulate the long-term gas migration in the buffer zone and to evaluate the feasibility of the buffer material, controlling the radionuclide-gas migration in the SNF repository site.

Spent nuclear fuels (SNFs) are stored in nuclear power plants for a certain period of time and then transported to an interim storage facility. After that, SNFs are finally repackaged in a disposal canister at an encapsulation plant for final disposal. Finland and Sweden have already completed the design of the spent nuclear fuel encapsulation plant. In particular, Finland has begun the construction of the encapsulation plant and is on the verge of completion. Korea Radioactive Waste Agency (KORAD) is conducting a conceptual design of a deep geological repository for SNFs. Conceptual design of the encapsulation plant is part of the research activity. It is highly required to draft an operation process of the encapsulation plant before an actual design activity. As part of the activity, Finnish design concept of the encapsulation plant and experience were thoroughly reviewed. Finally a preliminary concept of the operation process was proposed considering Korean unique situations such as the volume of SNFs estimated to be disposed of, types of transportation cask and other considerations.

APro, developed in KAERI for the process-based total system performance assessment (TSPA) of deep geological disposal systems, performs finite element method (FEM)-based multiphysics analysis. In the FEM-based analysis, the mesh element quality influences the numerical solution accuracy, memory requirement, and computation time. Therefore, an appropriate mesh structure should be constructed before the mesh stability analysis to achieve an accurate and efficient process-based TSPA. A generic reference case of DECOVALEX-2023 Task F, which has been proposed for simulating stationary groundwater flow and time-dependent conservative transport of two tracers, was used in this study for mesh stability analysis. The relative differences in tracer concentration varying mesh structures were determined by comparing with the results for the finest mesh structure. For calculation efficiency, the memory requirements and computation time were compared. Based on the mesh stability analysis, an approach based on adaptive mesh refinement was developed to resolve the error in the early stage of the simulation time-period. It was observed that the relative difference in the tracer concentration significantly decreased with high calculation efficiency.

In the design of HLW repositories, it is important to confirm the performance and safety of buffer materials at high temperatures. Most existing models for predicting hydraulic conductivity of bentonite buffer materials have been derived using the results of tests conducted below 100°C. However, they cannot be applied to temperatures above 100°C. This study suggests a prediction model for the hydraulic conductivity of bentonite buffer materials, valid at temperatures between 100°C and 125°C, based on different test results and values reported in literature. Among several factors, dry density and temperature were the most relevant to hydraulic conductivity and were used as important independent variables for the prediction model. The effect of temperature, which positively correlates with hydraulic conductivity, was greater than that of dry density, which negatively correlates with hydraulic conductivity. Finally, to enhance the prediction accuracy, a new parameter reflecting the effect of dry density and temperature was proposed and included in the final prediction model. Compared to the existing model, the predicted result of the final suggested model was closer to the measured values.

With the recent concern regarding cellulose enhancing radionuclide mobility upon its degradation to ISA, disposal of cellulosic wastes is being held off until the disposal safety is vindicated. Thus, a rational assessment should be conducted, applying an appropriate cellulose degradation model considering the disposal environment and cellulose degradation mechanisms. In this paper cellulose degradation mechanisms and the disposal environment are studied to propose the best-suitable cellulose degradation model for the domestic 1st phase repository. For the cellulose to readily degrade, the pH should be greater than 12.5. As in the case of SKB, 1BLA is excluded from the safety assessment because the pH of 1BLA remains below 12.5. Furthermore, despite cellulose degradation occurring, it does not always produce ISA. At low Ca2+ concentration, the ISA yield rate is around 25%, but at high Ca2+ concentration, the ISA yield rate increases up to 90%. Thus, for the cellulose to be a major concern, both pH and Ca2+ concentration conditions must be satisfied. To satisfy both conditions, the cement hydration must be in 2nd phase, when the porewater pH remains around 12.5 and a significant amount of Ca2+ ion is leaching out from the cement. However, according to the safety evaluation and domestic research, 2nd phase of cement hydration for silo concrete would achieve a pH of around 12.4, dissatisfying cellulose degradation condition like in 1BLA. Thus, cellulose degradation would be unlikely to occur in the domestic 1st phase repository. To derive waste acceptance criteria, a quantitative evaluation should be conducted, conservatively assuming cellulose is degraded. To conduct a safety evaluation, an appropriate degradation model should be applied to determine the degradation rate of cellulose. According to overseas research, despite the mid-chain scission being yet to be seen in the experiments, the degradation model considering mid-chain scission is applied, resulting in an almost 100% degradation rate. The model is selected because the repositories are backfilled with cement, achieving a pH greater than 13, so extensive degradation is reasonably conservative. However, under the domestic disposal condition, where cellulose degradation is unlikely to occur, applying such model would be excessively conservative. Thus, the peeling and stopping model derived by Van Loon and Haas, which suggests 10~25% degradation rate, is reasonably conservative. Based on this model, cellulose would not be a major concern in the domestic 1st phase repository. In the future, this study could be used as fundamental data for planning waste acceptance criteria.

In underground repository environments, various types of engineered barriers are installed to hinder the mobility of radionuclides. Cement admixtures, especially used to improve workability for concrete, are composed of fairly high organic molecules and have a dispersing effect through bonding with the C-S-H of the concrete. Previous studies have shown that complex-forming organics like EDTA, NTA, and ISA have a significant effect on the mobility of radionuclides, but the studies on the behavior and stability of combined complexes in hydrated cement are lacking. So, we selected a commonly used polycarboxylic-ester (PCE) type cement admixture and stable Co as a surrogate of Co-60 to perform desorption experiments from hydrated cement containing the admixture. Radioactive Co is known to be a common contaminant in nuclear fission and medical facilities and considered to exist as a relatively stable phase in repositories. In addition, the evaluation of cobalt can be a standard of safety issue for other radionuclides with the presence of cement admixture in repository. In this study, cement samples were prepared at water/cement ratio of 0.55 and cured for 28 days at 23-25°C and at least 80% of humidity with varying cement admixtures of 0.0, 0.1, and 2.0wt%. To evaluate the stability of cobalt in the weathered cement, a 0.001 M HCl solution was used to simulate cement weathering conditions on a hot plate at 60°C for 1 day using a solid/liquid ratio of 1:100. Degree of weathering was confirmed using XRD analysis. The adsorption experiments were performed by adding 0.0042 mmol of cobalt (CoCl2, Sigma-Aldrich, anhydrous ≥ 98.0%) to the weathered cement for 3 days using a platform shaker at 200 rpm, and the supernatant was separated using a syringe filter (<0.20 um) before ICP-MS analysis to determine the amount of Co adsorption. Cobalt desorption was tested for the Co-adsorbed cement using 0.019 mmol of calcium (Ca(NO3)2·4H2O, Sigma-Aldrich, 99%) for 3 hours to 14 days. The results showed that adsorbed cobalt with and without cement admixture was stably bound to cement, and did not increase any noticeable Co release by 2.0wt% PCE admixture. However, additional experiments using varying contents of PCE and other admixtures should be conducted to provide a standard for assessing the safety of cement admixtures in repositories.

The engineered barrier system (EBS) for deep geological disposal of high-level radioactive waste requires a buffer material that can prevent groundwater infiltration, protect the canister, dissipate decay heat effectively, and delay the transport of radioactive materials. To meet those stringent performance criteria, the buffer material is prepared as a compacted block with high-density using various press methods. However, crack and degradation induced by stress relaxation and moisture changes in the compacted bentonite blocks, which are manufactured according to the geometry of the disposal hole, can critically affect the performance of the buffer. Therefore, it is imperative to develop an adequate method for quality assessment of the compacted buffer block. Recently, several non-destructive testing methods, including elastic wave measurement technology, have been attempted to evaluate the quality and aging of various construction materials. In this study, we have evaluated the compressive wave velocity of compacted bentonite blocks via the ultrasonic velocity method (UVM) and free-free resonant column method (FFRC), and analyzed the relationship among compressive wave velocity, dry density, thermal conductivity, and strength parameter. We prepared compacted bentonite block specimens using the cold isostatic pressure (CIP) method under different water content and CIP pressure conditions. Based on multiple regression analysis, we suggest a prediction model for dry density in terms of manufacturing conditions. Additionally, we propose an empirical model to predict thermal conductivity and unconfined compressive strength based on compressive wave velocity. The database and suggested models in this study can contribute to the development of quality assessment and prediction techniques for compacted buffer blocks used in the construction of a disposal repository.

To conduct numerical simulation of a disposal repository of the spent nuclear fuel, it is necessary to numerically simulate the entire domain, which is composed on numerous finite elements, for at least several tens of thousands of years. This approach presents a significant computational challenge, as obtaining solutions through the numerical simulation for entire domain is not a straightforward task. To overcome this challenge, this study presents the process of producing the training data set required for developing the machine learning based hybrid solver. The hybrid solver is designed to correct results of the numerical simulation composed of coarse elements to the finer elements which derive more accurate and precise results. When the machine learning based hybrid solver is used, it is expected to have a computational efficiency more than 10 times higher than the numerical simulation composed of fine elements with similar accuracy. This study aims to investigate the usefulness of generating the training data set required for the development of the hybrid solver for disposal repository. The development of the hybrid solver will provide a more efficient and effective approach for analyzing disposal repository, which will be of great importance for ensuring the safe and effective disposal of the spent nuclear fuel.

For the performance analysis of deep geological repository systems, numerical simulation with multi-physics is required, which specifically covers Thermal (T), Hydraulic (H), and Mechanical (M) behaviors in the disposal environment. Numerous simulation models have been developed so far, each of which varies in the approach and methodology for solving THM problems. Fully-coupled THM simulation codes such as ROCMAS, THAMES, and CODE_BRIGHT were mainly developed in the initial stage of DEvelopment of COupled models and their VALidation against EXperiments (DECOVALEX), with the advantage of thorough calculations consisting of correlated several variables on different physics. Due to the difficulty of solving the complex Jacobian Matrix and the following burden for the computational calculation, weakly-coupled THM models have been suggested in recent researches: TOUGH2-MP with FLAC3D, TOUGH2 with UDEC and OpenGeoSys with FLAC3D. This methodology of loose coupling allows the practical use of computational code optimized for each physics, thereby increasing the efficiency in simulation. However, these suggested models require two different numerical codes to calculate THM behaviors, which leads to several inherent issues: compatibility during maintenance, updating and dependency between two codes. In this study, therefore, the authors build a unified code for simulating THM behaviors in the deep geological repository. The concept involves the iterative sequential coupling between TH and M for calculation efficiency. As having developed the simulation code, High-level rAdiowaste Disposal Evaluation System (HADES), to describe TH behavior based on Multi-physics Object-Oriented Simulation Environment (MOOSE) software, the authors make a milestone to develop and couple the MOOSE-based new code for M behavior as Sub-app, with the previous HADES set to be Main-app. New model for M behavior will be verified with the benchmark case of DECOVALEX-THMC Task D, comparing the mechanical simulation results: stress evolution over time, profiles of stress and vertical displacement. The existing simulation results from HADES will also be updated with the coupled calculations, with regard to temperature and saturation. Additionally, the effective stress evolution can be assessed in terms of repository’s stability with Spalling Strength and Mohr-Coulomb failure criterion. This concept for new simulation model has its meaning in that it aims to demonstrate the specific methodology of loosely coupling multi-physics in unified simulation code and analyze THM complex interactions with considering mutual influence on various physics. It is expected that HADES can be renewed as an integral simulation model for deep geological repository systems by possessing the capacity for analyzing and assessing mechanical behavior.

The high-level nuclear waste (HLW) repository is a 500-1,000 m deep underground structure to dispose high-level nuclear waste. The waste has a very long half-time and is exposed to a number of stresses, including high temperatures, high humidity, high pressure These stresses cause the structure to deteriorate and create cracks. Therefore, structural health monitoring with monitoring sensors is required for safety. However, sensors could also fail due to the stresses, especially high temperature. Given that the sensors are installed in the bentonite buffer and the backfill tunnel, it is impossible to replace them if they fail. That’s why it is necessary to assess the sensors’ durability under the repository’s environmental conditions before installing them. Accelerated life test (ALT) can be used to assess durability or life of the sensors, and it is important to obtain the same failure mode for reliability tests including ALT. Before conducting the test, the proper stress level must be designed first to get reliable data in a short time. After that, acceleration of life reduction with increasing temperature and temperature-life model should be determined with some statistical methods. In this study, a methodology for designing stress levels and predicting the life of the sensor were described.

Chemical environments of near-field (Engineered barrier and surrounded host rock) can influence performance of a deep geological repository. The chemical environments of near-field change as time evolves eventually reaching a steady state. During the construction of a deep geological repository, O2 will be introduced to the deep geological repository. The O2 can cause corrosion of Cu canisters, and it is important predicting remaining O2 concentration in the near-field. The remaining O2 concentration in the near field can be governed by the following two reactions: oxidation of Cu(I) from oxidation of Cu and oxidation of pyrite in bentonite and backfill materials. These oxidation reactions (Cu(I) and pyrite oxidation) can influence the performance of the deep geological repository in two ways; the first way is consuming oxidizing agents (O2) and the second way is the changing pH in the near-field and ultimately influencing on the mass transport rate of radionuclides from spent nuclear fuel (failure of canisters) to out of the engineered barrier. Hence, it is very important to know the evolution of chemical environments of near-field by the oxidation of pyrite and Cu. However, the oxidation kinetics of pyrite and Cu are different in the order of 1E7 which means the overall kinetics cannot be fully considered in the deep geological repository. Therefore, it is important to develop a simplified Cu and pyrite oxidation kinetics model based on deep geological repository conditions. Herein, eight oxidation reactions for the chemical species Cu(I) were considered to extract a simplified kinetic equation. Also, a simplified kinetics equation was used for pyrite oxidation. For future analysis, simplified chemical reactions should be combined with a Multiphysics Cu corrosion model to predict the overall lifetime of Cu canisters.

Since 1992, various numerical codes, such as TOUGH-FLAC and ROCMAS, have been developed and validated to dispose of Spent Nuclear Fuel (SNF) safely through a series of DEvelopment of COupled models and their VALidation against EXperiments (DECOVALEX) projects. These codes have been developed using different approaches, such as general two-phase flow and Richards’ flow which is an approximated approach neglecting gas pressure change, to implement the same multiphysics behaviors. However, the quantitative analysis for numerical results, which originated from different fundamental approaches, has not been conducted accurately. As a result, improper utilization of the approach to analyze certain conditions occurring such as dramatic gas pressure change may result in erroneous outcomes and systemic problem pertaining to TH analysis. In this study, the quantitative analysis of the two approaches, in terms of TH behavior, was conducted by comparing them with a 1D simulation of the CTF1 experiment carried out by laboratory experiment. The results calculated by different approaches show agreement in terms of TH behaviors and material properties change until 120°C. The results verify the applicability of Richards’ flow approach in a high temperature environment above the current thermal criteria, set as 100°C, and gas pressure change does not have a significant impact until 120°C. Therefore, although further studies for applicability of Richards’ flow are needed to suggest the appropriate temperature range, these quantitative analyses may contribute to the performance assessment of a compact repository using the high-temperature bentonite concept, which is currently gaining attention.

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.

The mobility of uranium (U) in the environment of a deep geological repository is controlled by various geochemical conditions and parameters. In particular, oxidation state of uranium is considered as a major factor to control the mobility of uranium in most of geological environments. In this study, therefore, we investigated the mobility of uranium in a deep geological repository by a natural analogue approach using a uranium deposit in the Ogcheon Metamorphic Belt (OMB). Uranium contents of rock samples from the study site ranged from 1.3 to 71 ppm (average 17.4 ppm). Uranium minerals found in the study site were mostly uraninite (UIVO2+x) and uranothorite ((UIV, Th)SiO4). The concentrations of U in the groundwater samples were very low (0.025~0.690 ppb) even though redox conditions are weakly oxidizing. Calculation results for U speciation in groundwater samples showed that major dissolved uranium species in the groundwater samples are mainly as calcium uranyl (UO2 2+) carbonate complexes such as Ca2UO2(CO3)3(aq) and CaUO2(CO3)3 2-. However, the activity ratios between 234U and 238U (AR(234U/238U)) showed U behavior in reducing conditions although the groundwater conditions were not reducing conditions and major dissolved U species were U(VI) species. Results from electron microscopic analyses for rock samples showed that major uranium minerals were U(IV) minerals such as uraninite and uranothorite. We could not identify other uranyl minerals and altered minerals from uraninite. This means that the geochemical condition of the study site has been maintained a reducing condition although the groundwater condition was a weakly oxidizing condition. Thus, the dissolution of uranium is strongly limited by the low solubility of uraninite. It is not obvious how the reducing condition of the study site has been maintained. Reducing agents such as pyrite, organic materials, and reducing bacteria might contribute to maintaining the reducing condition although further studies will be necessary. Results from this study imply that uranium mobility will be greatly limited by low dissolution of uraninite into groundwater if the reducing condition is well reserved. This limited mobility of uranium will be also contributed by low possibility of uraninite alteration into uranyl minerals which have a higher solubility than uraninite.

The nuclear criticality analyses considering burnup credit were performed for a spent nuclear fuel (SNF) disposal cell consisting of bentonite buffer and two different types of SNF disposal canister: the KBS-3 canister and small standardized transportation, aging and disposal (STAD) canister. Firstly, the KBS-3 & STAD canister containing four SNFs of the initial enrichment of 4.0wt% 235U and discharge burnup of 45,000 MWD/MTU were modelled. The keff values for the cooling times of 40, 50, and 60 years of SNFs were calculated to be 0.79108, 0.78803, and 0.78484 & 0.76149, 0.75683, and 0.75444, respectively. Secondly, the KBS-3 & STAD canister with four SNFs of 4.5wt% and 55,000 MWD/MTU were modelled. The keff values for the cooling times of 40, 50, and 60 years were 0.78067, 0.77581, and 0.77335 & 0.75024, 0.74647, and 0.74420, respectively. Therefore, all cases met the performance criterion with respect to the keff value, 0.95. The STAD canister had the lower keff values than KBS-3. The neutron absorber plates in the STAD canister significantly affected the reduction in keff values although the distance among the SNFs in the STAD canister was considerably shorter than that in the KBS-3 canister.

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.

In Korea, 483,102 assemblies of spent fuel have been discharged and stored in sites, as of 2019. However, total capacity for site storage is 529,748 assemblies, and more than 90% is already saturated. Wolsong site, the most saturated site, started to construct more dry storage to extend the capacity in 2020. Spent fuel and high-level waste (HLW) is a big concern in Korean nuclear industry. Then, master plan for management of spent fuel is once announced by Ministry of Trade, Industry and Energy (MOTIE) in 2016 and reviewed by civil committee in 2019. The core contents of the plan are establishing schedule for construction of HLW management facility in one area, and construction of temporary dry storage in each site, if unavoidable. For HLW management facility, there are three following schedules: siting of Underground Research Laboratory (URL) and Interim Storage by 2020, operation of facilities initiated by 2030, and operation of final disposal facility initiated by 2050. Final repository will be designed as deep geological repository. The concept of deep geological disposal is that spent nuclear fuel is placed in disposal containers that can withstand corrosion and pressure in long-term, permanently isolated from the human sphere of life, and dumped in deep geological media, such as crystalline rocks and clay layer, at a depth of 300 to 1,000 meters underground. The safety assessment of waste disposal sites focuses on determining whether the disposal sites meet the safety requirements of national regulatory authority. This safety assessment evaluates the potential radiation dose of radionuclides from the disposal site to humans or the environment. In this case, the calculation is performed assuming that all engineering barriers of the disposal site have collapsed in a long-term period. Then radionuclides are released from the waste, and migrated in groundwater. The dose resulting from the release and migration of radionuclides on the concentration of nuclides in groundwater. In general, metallic nuclides may exist in water in various ionic states, but some form colloids. This colloid allows more nuclides to exist in water than in solubility. Therefore, more doses may occur than we know generally predict. To determine the impact of colloids, we performed the safety assessment of the Yucca Mountain repository as an example.

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.

Glass wool, the primary material of insulation, is composed of glass fibers and is used to insulate the temperature of steam generators and pipes in nuclear power plants. Glass fiber is widely adopted as a substitute for asbestos classified as a carcinogen. The insulations used in nuclear power plants are classified as radioactive waste and most of the insulation is Very Low-Level Waste (VLLW). It is packaged in a 200 L drum the same as a Dry Active Waste (DAW). In the case of the insulations, it is packaged in a vinyl bag and then charged into the drum for securing additional safety because of the fine particle size of the fiberglass. A safety assessment of the disposal facility should be considered to dispose of radioactive waste. As a result of analyzing overseas Waste Acceptance Criteria (WAC), there is no case that has a separate limitation for glass fiber. Also, in order to confirm that glass fibers can be treated in the same manner as DAW, research related to the diffusion of glass fibers into the environment was conducted in this paper. It was confirmed that the glass fiber was precipitated due to the low flow velocity of groundwater in the Gyeongju radioactive waste repository and did not spread to the surrounding environment due to the effect of the engineering barrier. Therefore, the glass fiber has no special issue and can be treated in the same way as a DAW. In addition, it can be disposed of in the disposal facility by securing sufficient radiological safety as VLLW.