Metakaolin-based geopolymers have shown promise as suitable candidates for 14C immobilization and final disposal. It has been shown that the physicochemical properties of metakaolin wasteforms meet, and often far exceeding, the strict compression strength and leaching acceptance criteria of the South Korea radioactive waste disposal site. However, it is not possible to analyze and characterize the internal structure of the geopolymer wasteform by conventional characterization techniques such as microscopy without destruction of the wasteform; an impractical solution for inspecting wasteforms destined for final disposal. Internal inspection is important for ensuring wastes are homogenously mixed throughout the wasteform and that the wasteform itself does not pose any significant defects that may have formed either during formulation and curing or as a result of testing prior to final disposal. X-ray Computed Tomography (XCT) enables Non-Destructive Evaluation (NDE) of objects, such as final wasteforms, allowing for both their internal and external, characterization without destruction. However, for accurate quantification of an objects dimensions the spatial resolution (length and volume measures) must be know to a high degree of precision and accuracy. This often requires extensive knowledge of the equipment being used, its precise set-up, maintenance and calibration, as well as expert operation to yield the best results. A spatial resolution target consists of manufactured defects of uniformed dimensions and geometries which can be measured to a high degree of accuracy. Implementing the use of a spatial resolution target, the dimensions of which are known and certified independently, would allow for rapid dimensional calibration of XCT systems for the purpose of object metrology. However, for a spatial resolution target to be practical it should be made of the same material as the intended specimen, or at least exhibit comparable X-ray attenuation. In this study, attempts have been made to manufacture spatial resolution targets using geopolymer, silica glass, and alumina rods, as well as 3D printed materials with varying degrees of success. The metakaolin was activated by an alkaline activator KOH to from a geopolymer paste that was moulded into a cylinder (Diameter approx. 25 mm). The solidified geopolymer cylinder as well as both the silica glass rod and alumina rod (Diameter approx. 25 mm) we cut to approximately 4 mm ± 0.5 mm height with additional end caps cut measuring 17.5 mm ± 2.5 mm height. All parts were then polished to a high finish and visually inspected for their suitability as spatial resolution targets.

The decommissioning of Korea Research Reactor Units 1 and 2 (KRR-1&2), the first research reactors in South Korea, began in 1997. Approximately 5,000 tons of waste will be generated when the contaminated buildings are demolished. Various types of radioactive waste are generated in large quantities during the operation and decommissioning of nuclear facilities, and in order to dispose of them in a disposal facility, it is necessary to physico-chemically characterize the radioactive waste. The need to transparently and clearly conduct and manage radioactive waste characterization methods and results in accordance with relevant laws, regulations, acceptance standards is emerging. For radioactive waste characterization information, all information must be provided to the disposal facility by measuring and testing the physical, chemical, and radiological characteristics and inputting related documents. At this time, field workers have the inconvenience of performing computerized work after manually inputting radioactive waste characterization information, and there is always a possibility that human errors may occur during manual input. Furthermore, when disposing of radioactive waste, the production of the documents necessary for disposal is also done manually, resulting in the aforementioned human error and very low production efficiency of numerous documents. In addition, as quality control is applied to the entire process from generation to treatment and disposal of radioactive waste, it is necessary to physically protect data and investigate data quality in order to manage the history information of radioactive waste produced in computerized work. In this study, we develop a system that can directly compute the radioactive waste characterization information at the field site where the test and measurement are performed, protect the stored radioactive waste characterization data, and provide a system that can secure reliability.

Most of the spent nuclear fuel generated by domestic nuclear power plants (NPPs) is temporarily stored in wet storage which is spent fuel pool (SFP) at each site. Currently, in case of Kori Unit 2, about 93.6% of spent nuclear fuel is stored in SFP. Without clear disposal policy determined for spent nuclear fuel, the storage capacity in each nuclear power plant is expected to reach saturation within 2030. Currently, the SFP stores not only spent fuel but also various non-fuel assembly (NFA). NFA apply to all device and structures except for fuel rods inserted in nuclear fuel assembly. The representative NFA is control element driving mechanism (CEDM), in-core instrument (ICI), burnable poison, and neutral resources. Although these components are irradiated in the reactor, they do not emit high-temperature heat and high radiation like nuclear fuel, so if they are classified as intermediate level waste (ILW) and low level waste (LLW) and moved outside the SFP, positive effects such as securing spent fuel storage space and delaying saturation points can be obtained. Therefore, this study analyzes the status of spent fuel and Non Fuel Assembly (NFA) storage in SFP of domestic nuclear power plants. In addition, this study predict the amount of spent fuel and NFA that occur in the future. For example, this study predicts the percentage of current and future ICIs and control rods in the SFP when stored in the spent fuel storage rack. In addition, the positive effects of moving NFA outside the SFP is analyzed. In addition, NFA withdrawn from SFP is classified as ILW & LLW according to the classification criteria, and the treatment, storage, and disposal methods of NFA will be considered. The study on the treatment, storage, and disposal methods of NFA is planned to be conducted by applying the existing KN-12 & KN-18 containers and ILW & LLW containers being developed for decommissioning waste.

The spent filters used to purify radioactive materials and remove impurities from primary systems at nuclear power plants (NPPs) have been stored for long periods in filter storage rooms at NPPs due to concerns about the unproven safety of the treatment method, absence of disposal facilities, and risk of high radiation exposure. In the storage room at Kori Unit 1, there are approximately 227 spent filters of 9 different types. The radiation dose rates of filters range from 0.01 to 500 mSv/hr. Recently, a comprehensive plan has been established for the treatment and disposal of radioactive waste that has not yet been treated to facilitate decommissioning of NPPs. As a follow-up measure, compression and packaging optimization processes are being developed to treat the spent filters. KHNP plans to dispose of the spent filters after compressing, packaging, and immobilizing them. However, the spent filters are currently stored without being sorted by type or radiation intensity. If the removal and packing of the filters are done randomly without a plan for the order of withdrawal and subsequent processes, issues may arise such as a decrease in drum loading efficiency and exceeding the dose limit of the package. In this study, the number of drums needed to pack the spent filters was calculated, considering the filter size, weight, quantity, dose rate, shielding thickness of drum, and loadable quantity in a shielding drum (SD). Then, the spent filters that can be loaded on each drum were classified into one group. In addition, the withdrawal order for each group was set so that the filter withdrawal, compression, and packaging processes could be performed efficiently. The spent filter groups are as follows: (1) compression/12 cm SD (17 groups), (2) compression/16 cm SD (6 groups), (3) non-compression/ intermediate storage container (17 groups, additional radiation attenuation required due to high dose rate), and (4) unclassified (5 groups, determined after measurement due to lack of filter information). The withdrawal order of the groups was determined based on several factors, including visual identification of the filter, ease of distribution after withdrawal, work convenience, and safety. Due to the decay of radioactivity over time, the current dose rate of the spent filters is expected to be much lower than at the time of waste generation. Therefore, in the future, sample filters will be taken from the storage room to measure their radioactivity and radiation dose rate. Based on these measurements, a database of radiological characteristics for the 227 filters will be created and used to revise the filter grouping.

Bentonite is a widely used buffer material in high-level radioactive waste repositories due to its favorable properties, including its ability to swell and low permeability. Bentonite buffers play an important role in safe disposal by providing a low permeability barrier and preventing radionuclides migration into the surrounding rock. However, the long-term performance of the bentonite buffer is still an area of research, and one of the main concerns is the erosion of the buffer due to swelling and groundwater flow. Erosion of the bentonite buffer can have a significant impact on repository safety by reducing the integrity of the buffer and forming colloids that can transport radionuclides through groundwater, potentially increasing the risk of radionuclide migration. Therefore, understanding the mechanisms and factors that influence the erosion of the bentonite buffer is critical to the safety assessment of high-level radioactive waste repositories. In this study, we attempted to develop the bentonite buffer erosion model using Adaptive Processbased total system performance assessment framework for a geological disposal system (APro) proposed by the Korea Atomic Energy Research Institute (KAERI). First, the erosion phenomenon was divided into two stages: bentonite buffer penetration into rock fractures and colloid formation. As an initial step in the development of the buffer erosion model, a bentonite buffer intrusion into the fracture and consequent degradation of buffer property were considered. For this purpose, a tworegion model based on the dynamic bentonite diffusion model was adopted which is one of the methods for simulating bentonite buffer intrusion. And, it was assumed that the buffer properties, such as density, porosity and permeability, thermal conductivity, modulus of elasticity, and mechanical strength, are degraded as the buffer erodes. The bentonite buffer degradation model developed in this study will serve as a foundation for the comprehensive buffer erosion model, in conjunction with the colloidal formation model in the future.

It is important to make a strategy for clearance-level radioactive waste. Sampling and disposal plans should be drawn up with characteristics of target waste. In this paper, a target clearance-level radioactive waste is used in a laboratory for experiments with Cs-137 and Co-60, unsealed radioactive sources with gamma radiation isotopes. Therefore, it is enough to analyze with HPGe to check the contaminant level. The laboratory fume hood combined multiple materials, which means some are volume contamination and others are surface contamination. The wood, plastic, and drywall boards, which are absorbent volume contaminated parts and make up PVC pipes, base cabinet doors, backside baffles, etc., will be sampled with coring methods. The metals and glasses, which are unabsorbent, surface-contaminated parts, are sampled with smear methods. The work surface, baffles, exhaust plenum, and glass sash inside parts have a high possibility of being contaminated. The hood body, flame, base cabinet, PVC pipe (the rare end of the filter), and blower transition case have a low possibility of becoming contaminated. When we checked with HPGe, except for the work surface (which was below clearance level), other parts were less than MDA. The highest radionuclide concentration was in PVC pipe: Cs-137C 3.91E-02 (Bq/g), Co-60 4.54E- 03 (Bq/g). It is less than clearance level. Therefore, the waste was applied for the clearance level radioactive wastes and got permission from the regulatory body.

The nuclear facilities at Korea Atomic Energy Research Institute (KAERI) have generated a variety of liquid radioactive waste and most of them have low-level radioactive or lower levels. Some of the liquid radioactive waste generated in KAERI is transported to Radioactive Waste Treatment Facility (RWTF) in 20 L container. Liquid radioactive waste transported in a 20 L container is stored in a Sewer Tank after passing through a solid-liquid separation filter. It is then transferred to a very low-level liquid radioactive waste Tank after removing impurities such as sludge through a pre-treatment device. The previous pre-treatment process involved an underwater pump and a cartridge filter device passively, but this presented challenges such as the inconvenience of having to install the underwater pump each time, radiation exposure for workers due to frequent replacement of the cartridge filter, and the generation of large amounts of radioactive waste from the filter. To address these challenges and improve efficiency and safety in radiation work, an automated liquid radioactive waste pre-treatment device was developed. The automated liquid radioactive waste pre-treatment device is a pressure filtration system that utilizes multiple overlapping filter plates and pump pressure to effectively remove impurities such as sludge from liquid radioactive waste. With just the push of a button, the device automatically supplies and processes the waste, reducing radiation hazards and ensuring worker safety. Its modular and mobile design allows for flexible utilization in various locations, enabling efficient pre-treatment of liquid radioactive waste. To evaluate the performance of the newly constructed automated liquid radioactive waste treatment device, samples were taken before and after treatment for 1 hour cycling and analyzed for turbidity. The results showed that the turbidity after treatment was more than about four times lower than before treatment, confirming the excellent performance of the device. Also, it is expected that the treatment efficiency will improve further as the treatment time and number of cycles increase.

Level measurement of liquid radwaste is essential for inventory management of treatment system. Among various methods, level measurement based on differential pressure has many advantages. First, it is possible to measure the liquid level of the system regardless of liquid type. Second, as the instrument doesn’t need to be installed near the tank, there is no need to contact the tank when managing it. Therefore, workers’ radiation dose from the system can be decreased. Finally, although it depends on the accuracy, the price of the instrument is relatively low. With these advantages, in general, liquid radwaste level in a tank is measured using differential pressure in the treatment system. Not only the advantages described above, there are some disadvantages. As the liquid in the system is waste, it is not pure but has some suspended materials. These materials can be accumulated in tanks and pipes where the liquids move to come into direct contact with pneumatic pipes that are essential in differential pressure instruments. As a result, in case of a treatment using heat source, the accumulated materials may become sludge causing interference in pneumatic pipes. And this can change the pressure which also affects the level measured. In conclusion, in case of liquid storage tanks in which the situation cannot be checked, the proficiency of an operator becomes important.

Surface water temperature of a bay (from the south to the north) increases in spring and summer, but decreases in autumn and winter. Due to shallow water depth, freshwater outflow, and weak current, the water temperature in the central to northern part of the bay is greatly affected by the land coast and air temperature, with large fluctuations. Water temperature variations are large in the north-east coast of the bay, but small in the south-west coast. The difference between water temperature and air temperature is greater in winter and in the south-central part of the bay than that in the north to the eastern coast of the bay where sea dykes are located. As the bay goes from south to north, the range of water temperature fluctuation and the phase show increases. When fresh water is released from the sea dike, the surrounding water temperature decreases and then rises, or rises and then falls. The first mode of empirical orthogonal function (EOF) represents seasonal variation of water temperature. The second mode represents the variability of water temperature gradient in east-west and north-south directions of the bay. In the first mode, the maximum and the minimum are shown in autumn and summer, respectively, consistent with seasonal distribution of surface water temperature variance. In the second mode, phases of the coast of Seosan~Boryeong and the east coast of Anmyeon Island are opposite to each other, bordering the center of the deep bay. Periodic fluctuation of the first mode time coefficient dominates in the one-day and half-day cycle. Its daily fluctuation pattern is similar to air temperature variation. Sea conditions and topographical characteristics excluding air temperature are factors contributing to the variation of the second mode time coefficient.

Some of the metal waste generated from KEPCO NF is being disposed of in the form of ingots. An ingot is a metal that is melted once and then poured into a mold to harden, and it is characterized by a uniform distribution of radioactive material. When measuring the uranium radioactivity in metal ingot with HPGe detector, 185.7 keV of U-235 is used typically because most gamma rays emitted at U-235 are distributed in low-energy regions below 200 keV. To analyze radioactivity concentration of U-235 with HPGe detector more accurately, self-attenuation due to geometrical differences between the calibration source and the sample must be corrected. In this study, the MCNP code was used to simulate the HPGe gamma spectroscopy system, and various processes were performed to prove the correlation with the actual values. First an metal ingottype standard source was manufactured for efficiency calibration, and the GEB coefficient was derived using Origin program. And through the comparison of actual measurements and simulations, the thickness of the detector’s dead layers were defined in all directions of Ge crystal. Additionally instead of making an metal ingot-type standard source every time, we analyzed the measurement tendency between commercially available HPGe calibration source (Marinelli beaker type) and the sample (metal ingot type), and derived the correction factor for geometry differences. Lastly the correction factor was taken into consideration when obtaining the uranium radioactivity concentration in the metal ingot with HPGe gamma spectroscopy. In conclusion, the U-235 radioactivity in metal ingot was underestimated about 25% of content due to the self-attenuation. Therefore it is reasonable to reflect this correction factor in the calculation of U-235 radioactivity concentration.

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 prove the long-term safety of deep geological repository, the safety assessment is needed to ensure that the expected performance of repository satisfies the regulatory standards. Scenario development is process of analyzing events and evolutions that can directly or indirectly affect the performance of a disposal system and is a pre-step for quantitative safety assessment. Scenarios are used to identify and define cases to be assessed by numerical modeling, and cases are mainly divided into normal (also called the ‘reference’ and ‘expected evolution’) and abnormal scenarios. Mainly two approaches have been used to set up scenarios. One is a bottom-up approach that starts with features, events and processes (FEPs). This approach can analyze the evolution and events related to the performance of the disposal system in an inductive manner. The other is top-down approach that analyzes the events and evolution of disposal system, focusing on situations that may affect the safety function of the components. This approach starts with a set of intuitively predefined expected failures of safety function. Combining the two approaches is more effective in demonstrating comprehensiveness which is a main challenge of scenario analysis, and almost national radioactive waste management institutions combine top-down and bottom-up approaches for development of scenarios. An approach combining the two approaches is called a hybrid approach, and the detailed method differs from each institution and has not been determined. In this study, some work for constructing the scenario using hybrid approach was performed. Firstly, defining each component’s safety function and screening FEPs according to several rules were performed for a generic repository. Secondly, we extracted performance factors that are considered likely to affect safety functions. And lastly, we integrated FEPs correlated with performance factor to simplify the analysis. These results will be material to construct the scenario using hybrid approach.

The most important thing in development of a process-based TSPA (Total System Performance Assessment) tool for large-scale disposal systems (like APro) is to use efficient numerical analysis methods for the large-scale problems. When analyzing the borehole in which the most diverse physical phenomena occur in connection with each other, the finest mesh in the system is applied to increase the analysis accuracy. Since thousands of such boreholes would be placed in the future disposal system, the numerical analysis for the system becomes significantly slower, or even impossible due to the memory problem in cases. In this study, we propose a tractable approach, so called global-local iterative analysis method, to solve the large-scale process-based TSPA problem numerically. The global-local iterative analysis method goes through the following process: 1) By applying a coarse mesh to the borehole area the size of the problem of global domain (entire disposal system) is reduced and the numerical analysis is performed for the global domain. 2) Solutions in previous step are used as a boundary condition of the problem of local domain (a unit space containing one borehole and little part of rock), the fine mesh is applied to the borehole area, and the numerical analysis is performed for each local domain. 3) Solutions in previous step are used as boundary conditions of boreholes in the problem of global domain and the numerical analysis is performed for the global domain. 4) steps 2) and 3) are repeated. The solution derived by the global-local iterative analysis method is expected to be closer to the solution derived by the numerical analysis of the global problem applying the fine mesh to boreholes. In addition, since local problems become independent problems the parallel computing can be introduced to increase calculation efficiency. This study analyzes the numerical error of the globallocal iterative analysis method and evaluates the number of iterations in which the solution satisfies the convergence criteria. And increasing computational efficiency from the parallel computing using HPC system is also analyzed.

The timescale of safety assessment for a geological disposal system is considered up to hundreds of thousands of years when the radionuclides in spent nuclear fuel decay to levels comparable to natural radioactivity. During this long period, a variety of climate changes are expected to occur, including variations in temperature and precipitation as well as long-term sea level changes and glacial cycles. These climate changes can either directly affect water balance components or indirectly affect water balance by altering terrain and vegetation that have an impact on water balance. Water balance is a significant element of safety assessment, because it affects the radionuclide transport via groundwater flow, which in turn affects the radiological risk to humans and other biotas. Therefore, it is important to understand the hydrologic response to climate changes for proving the long-term safety of the disposal system. To this end, this study performed hydrological simulations using the SWAT (Soil and Water Assessment Tool) for several climate change scenarios. SWAT is the watershed-scale hydrological model developed by the USDA-ARS (United States Department of Agriculture - Agricultural Research Service) and has been widely used to quantify the water balance in a watershed. It calculates the hydrologic cycle based on the water balance equation with different physical processes for water balance components such as evapotranspiration, surface runoff, and groundwater recharge. This study assumed several climate change scenarios (e.g., variations in temperature and precipitation, sea level change, and formation of permafrost) and analyzed how the components of the water balance would respond under different scenarios and which scenarios would have the greatest impact on the water balance. These findings can provide valuable insights for future long-term safety assessments on the Korean Peninsula and can also be used as input data for the biosphere module of APro (Adaptive process-based total system performance assessment framework).

Since high-level radioactive wastes contain long-lived nuclides and emit high energy, they should be disposed of permanently through a deep geological disposal system. In Korea, the first (2016.07) and the second (2021.12) basic plans for the management of high-level disposal systems were proposed to select sites for deep geological disposal facilities and to implement business strategies. Leading countries such as Finland, Sweden and France have developed and applied safety cases to verify the safety of deep geological disposal systems. By examining the regulatory status of foreign leading countries, we analyze the safety cases ranging from the site selection stage of the deep geological disposal system to the securing of the permanent disposal system to the investigation, analysis, evaluation, design, construction, operation, and closure. Based on this analysis, we will develop safety case elements for long-term safety of deep geological disposal systems suitable for domestic situation. To systemically analyze data based on safety cases, we have established a database of deep geological disposal system regulations in leading foreign countries. Artificial intelligence text mining and data visualization techniques are used to provide database in dashboard form rather than simple lists of data items, which is a limitation of existing methods. This allows regulatory developers to understand information more quickly and intuitively and provide a convenient interface so that anyone can easily access the analyzed data and create meaningful information. Furthermore, based on the accumulated bigdata, the artificial intelligence learns and analyzes the information in the database through deep learning, and aims to derive a more accurate safety case. Based on these technologies, this study analyzed the legal systems, regulatory standards, and cases of major international leading countries and international organizations such as the United States, Sweden, Finland, Canada, Switzerland, and the IAEA to establish a database management system. To establish a safety regulation base suitable for the domestic deep geological disposal environment, the database is provided as data to refer to and apply systematic information management on regulatory standards and regulatory cases of overseas leading countries, and it is expected that it will play a key role as a forum for understanding and discussing the level of safety of deep geological disposal system among stakeholders.

The engineered barrier system (EBS) is an indispensable element of a deep geological repository (DGR) designed to prevent the discharge of radioactive materials into the environment. The buffer material is a vital component of the EBS by creating a physical and chemical barrier that prevents the migration of radioactive materials. In the disposal environment, gases can be generated from the corrosion of the canister. When the gas generation rate exceeds the diffusion rate, the buffer material’s performance can deteriorate by the physical damage induced by the increase in pore pressure. Therefore, understanding the EBS’s behavior under gas generation conditions is crucial to guarantee the longterm safety and performance of the DGR. Lab-scale and field-scale experiments have been conducted to examine the stability of the buffer material concerning gas generation and movement by the previous researchers. To evaluate long-term stability for more than 100,000 years, it is essential to assess stability using a numerical model verified by these experiments. This study investigated the effect of interfacial characteristics on the numerical modeling accuracy of experimental simulation while verifying a numerical model through field-scale experimental results. The findings of this study are expected to furnish fundamental data for establishing numerical analysis guidelines for the longterm stability assessment of disposal systems.

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.

With the increasing demand for a repository to safely dispose of high-level radioactive waste (HLW), it is imperative to conduct a safety assessment for HLW disposal facilities for ensuring the permanent isolation of radionuclides. For this purpose, the Korea Atomic Energy Research Institute (KAERI) is currently developing the Adaptive Process-based total system performance assessment framework for a geological disposal system (APro). A far-field module, which specifically focuses on fluid flow and radionuclide transport in the host rock, is one of several modules comprising APro. In Korea, crystalline rock is considered the host rock for deep geological disposal facilities due to its high thermal conductivity and extremely low permeability. However, the presence of complex fracture system in crystalline rock poses a significant challenge for managing fluid flow and nuclide transport. To address this challenge, KAERI is participating in DECOVALEX-2023 Task F1, which seeks to compare and verify modeling results using various levels of performance assessment models developed by each country for reference disposal systems. Through the benchmark problems suggested by DECOVALEX-2023 Task F1, KAERI adopts the Discrete Fracture-Matrix (DFM) as the primary fracture modeling approach. In this study, the transport processes of reactive tracers in fractured rock, modeled with DFM, are simulated. Specifically, three different tracers (conservative, decaying, adsorbing) are introduced through the fracture under identical injecting conditions. Thereafter, the breakthrough curves of each tracer are compared to observe the impact of reactive tracers on nuclide transport. The results of this study will contribute to a better understanding of nuclide behavior in subsurface fractured rock under various conditions.

The development of Features, Events, and Processes (FEPs) and scenarios, which consider the longterm evolution of repository, is underway, along with the construction of input data and a model database for the adaptive process-based total system performance assessment framework, APro. PAPiRUS serves as an integrated information processing platform, enabling users to seamlessly access, search, and extract essential information. To enhance data usability, it is crucial to establish well-structured metadata for each dataset. Regarding FEPs, individual FEPs consist of extensive text-based data and sets of other short textual data. To enhance the searchability of these FEPs, precise keywords must be assigned to each FEP. For user convenience, the PAPiRUS FEP database contains several FEPs not only the long-term evolution FEPs developed by KAERI but also thousands of FEPs form the databases such as NEA PFEPs and Posiva FEPs. Generating keywords for thousands of FEPs proves to be a labor-intensive task. Consequently, this study explores natural language processing techniques for keyword analysis to boost the productivity of the keyword generation process. Specifically, we employ Generative Pretrained Transformer (GPT) models for keyword extraction. Our test results for keyword extraction demonstrate that, although not flawless, providing suitable prompts yields sufficiently useful keyword sets. We identified several optimal prompts and developed an Excel-based program to derive keywords from the existing FEP database using these prompts. By using the outcomes of this study, initial versions of keyword sets for thousands of FEPs can be rapidly produced and subsequently refined through expert review and editing. The generated keywords will serve as metadata within PAPiRUS.

The Korean Nuclear Safety and Security Commission has established a general guideline for the disposal of high-level waste, which requires that radiological effects from a disposal facility should not exceed the regulatory safety indicator, a radiological risk. The post-closure safety assessment of the disposal facility aims to evaluate the radiological dose against a representative person, taking into account nuclide transport and exposure pathways and their corresponding probabilities. The biosphere is a critical component of radiation protection in a disposal system, and the biosphere model is concerned with nuclide transport through the surface medium and the doses to human beings due to the contaminated surface environment. In past studies by the Korea Atomic Energy Research Institute (KAERI), the biosphere model was constructed using a representative illustration of surface topographies and groundwater conditions, assuming that the representative surface environment would not change in the future. Each topography was conceptualized as a single compartment, and distributed surface contamination over the geometrical domain was abstracted into 0D. As a result, the existing biosphere model had limitations, such as a lack of quantitative descriptions of various transport and exposure pathways, and an inability to consider the evolution of the surface environment over time. These limitations hinder the accurate evaluation of radiological dose in the safety assessment. To overcome these limitations, recent developments in biosphere modeling have incorporated the nuclide transport process over a 2D or 3D domain, integrating the time-dependent evolution of the surface environment. In this study, we reviewed the methodology for biosphere modeling to assess the radiological dose given by distributed surface contamination over a 2D domain. Based on this review, we discussed the model requirements for a numerical module for biosphere dose assessment that will be implemented in the APro platform, a performance assessment tool being developed by the KAERI. Finally, we proposed a conceptual model for the numerical module of dose assessment.