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.

A lot of solid wastes are generated when nuclear power plant is dismantled, and a lot of treatment costs and optimal waste treatment technologies are required to treat the generated solid wastes. Currently, there is no optimized reduction and solidification technology for each characteristics of radioactive dismantling waste, so the customized treatment technology for each waste is required to respond actively to this issue. This paper shows the evaluation results of molding and sintering characteristics using preliminary sample to derive operational characteristics and improvements for powder mixing device, molding device, and sintering device manufactured for solidification of dispersible radioactive waste. Zeolite was used as a preliminary sample for performing basic operation characteristics evaluation of each unit device. First of all, the basic operation characteristics of the powder mixing device was evaluated by analyzing the sample distribution, mixing degree, and tap density. It was confirmed that the preliminary sample was well mixed in all areas of the cylinder where the mixing was performed. In the tap density analysis, the increase effect of the volume reduction of the sample was confirmed according to the increase of the RPM speed (up to 2000 RPM). Since the particle size of zeolite sample is very small (nanometer size), the particular consistency of the change of average particle size with RPM speed couldn’t be confirmed, but the uniform of particle size distribution was confirmed with RPM speed size. The basic operation characteristics of the molding device was evaluated for each mold size (ID30, ID50, ID100) according to the moisture content (0-20%) and the molding pressure condition (25-200 MPa) for the preliminary sample. In the characteristics evaluation of the sintered body, the strength of the sintered body was much higher than that of the molded body. However, it was confirmed that as moisture evaporated during the sintering process according to the moisture content contained in the molded body, the swelling occurred in the sintered body due to vapor pressure, and this caused cracks in the longitudinal or transverse direction inside and outside the sintered body. Therefore, optimal moisture content conditions for sintering should be derived. In conclusion, if the operation characteristics and improvements of powder mixing, molding and sintering devices derived from this study are reflected and improved, it is judged that it is possible to derive the optimal process for solidification of dispersive radioactive wastes.

The homogeneity of radioactive spent ion exchange resins (IERs) distribution inside waste form is one of the important characteristics for acceptance of waste forms in long-term storage because heterogenous immobilization can lead to the poor structural stability of waste form. In this study, the homogeneity of metakaolin-based geopolymer waste form containing simulant IERs was evaluated using a laser-induced breakdown spectroscopy (LIBS) and statistical approach. The cation-anion mixed IERs (IRN150) were used to prepare the simulant spent IERs contaminated by non-radioactive Cs, Fe, Cr, Mn, Ni, Co, and Sr (0.44, 8.03, 6.22, 4.21, 4.66, 0.48, and 0.90 mg/g-dried IER, respectively). The K2SiO3 solution to metakaolin ratio was kept constant at 1.2 and spent IERs loading was 5wt%. For the synthesis of homogeneous geopolymer waste form, spent IERs were mixed with K2SiO3 solution and metakaolin first, and then the fresh mixture slurry was poured into plastic molds (diameter: 2.9 cm and height: 6.0 cm). The heterogeneous geopolymer waste form was also fabricated by stacking two kinds of mixtures (8wt% IERs loading in bottom and 2wt% in top) in one mold. Geopolymers were cured for 7d (1d at room temperature and 6d at 60°C). The hardened geopolymers were cut into top, middle, and bottom parts. The LIBS spectra and intensities for Cs were obtained from the top and bottom of each part. Cs was selected for target nuclide because of its good sensitivity for measurement. Shapiro-Wilk test was performed to determine the normality of LIBS data, and it revealed that data from the homogeneous sample is normal distribution (p-value = 0.9246, if p-value is higher than 0.05, it is considered as normal distribution). However, data from the heterogeneous sample showed abnormal distribution (p-value = 7.765×10-8). The coefficient of variation (CoV) was also calculated to examine the dispersion of data. It was 31.3% and 51.8% from homogeneous and heterogeneous samples, respectively. These results suggest that LIBS analysis and statistical approaches can be used to evaluate the homogeneity of waste forms for the acceptance criterion in repositories.

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.

Radioactive waste must be stored for at least 300 years and must bear astronomical costs. In addition, unexpected potential risk factors are also a considerable burden. In the case of low-level radioactive waste, combustible and liquid low-contamination radioactive waste can be treated relatively easily through high-temperature plasma which the volume can be reduced by 1/250 and the weight by 1/30. It is possible to permanently dispose of the ash leached after plasma treatment in a more stable manner compared to the conventional methods. Types of low-level combustible radioactive waste, including paper, vinyl, clothing, filters, and resins, account for more than 30% of the total waste volume. Furthermore, high-temperature plasma treatment of low-level radioactive waste from petrochemical plants and medical institutions have many advantages, namely astronomical cost savings, securing free space in existing storage facilities, and improving the image of nuclear energy. Korea is preparing to decommission the Kori No. 1 nuclear power plant, and small and mediumsized enterprises and related organizations are conducting various studies to incinerate radioactive waste. In foreign countries, Britain began incineration technology in the 1970s, and Plasma Energy Group, LLC, headquartered in Florida, USA, physically changed the molecular structure of the material by combining plasma chambers and plasma arcs and obtained a patent application in 1992. Germany was approved for operation in 2002, and Switzerland completed a trial run of a plasma technologybased facility in 2004. Important radionuclides in terms of radioactive gas waste treatment include inert gases, radioiodine, and radioactive suspended particles. Gas waste is compressed in a compressor through a surge tank in the gas waste treatment system and filters at each stage. after that, the shortlife nuclide is naturally collapsed for 30 to 60 days in the storage or activated carbon adsorbent in the attenuation tank and released through HEPA filters. The radioactive concentration at discharge is monitored and managed using continuous monitoring equipment, and the oxygen concentration is managed in the gas waste treatment system to prevent explosion risk. The problem of radioactive waste disposal is not only a problem for people living in the present era, but also a big social issue that brings a burden to future generations While interest in plasma treatment is increasing from the decommissioning of the Kori Unit 1. in Korea, it is showed that there is a lack of systematic management and research especially in the radioactive volatile gases fields, that’s why I propose some ideas as follows. First, the government and related institutions should invest to the continuous radioactive monitoring system to produce and distribute continuous radioactive monitoring facilities with an affordable price. Second, it is recommended that radioactive waste incineration would be connected to the GRS system of the plant’s gas radwaste treatment system, and radioactive volatile materials should be monitored through continuous monitoring system. Third, radioactive volatile materials generated according to the temperatures and times during plasma incineration treatment are different. Therefore, prior classification of each expected radioactive volatile substance must be performed before incineration.

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.

The treatment of waste generated during operation as a part of preparation for decommissioning is coming to the fore as a pending issue. Non-fuel waste stored in the spent fuel pool (SFP) of PWRs in Korea includes Dummy fuel, damaged fuel rod storage container, reactor vessel specimen cask, spent in-core instrumentation, spent control element assemblies, spent neutron source assemblies, burnable poison rods, etc. In order to treat such waste, it is necessary to classify radioactive waste level and analyze kinds of nuclide in accordance with legal requirements. In order to solve the problem, the items that KHNP-CRI is trying to conduct like followings. First, KHNP-CRI will identify the current status of non-fuel waste stored in the SFP of all domestic nuclear power plants. In order to consider the treatment of non-fuel waste, it is essential to know what kind of items and how many items are stored in the SFP. Second, to identify the dimension and characteristics of non-fuel waste stored in the SFP would be conducted. The configuration of non-fuel waste is important information to handle them. Third, the way to handle non-fuel waste would be deduced including analysis of their dimension, whether the equipment should be developed to handle each kind of non-fuel waste or not, how to transport them. In order to classify radioactive waste level and analyze the nuclide for the non-fuel waste, handling tools and the cask to transport them into the facility which nuclide analysis is able to be performed would be required. Fourth, the nuclide analysis technology would be identified. Also, domestic holding technology would be identified and which technology should be developed to classify the radioactive waste level for the non-fuel waste would be deduced. This preliminary study will provide KHNP-CRI with the insight for the nuclide analysis technology and future work which is following action for the non-fuel waste. Based on the result of above preliminary study, the feasibility of the research for the treatment of non-fuel waste would be evaluated and research plan would be established. In conclusion, the treatment of non-fuel waste stored in the spent fuel pool of domestic PWR should be considered to prepare the decommissioning. KHNP-CRI will identify the quantity, the dimension and kinds of non-fuel waste in the SFP of domestic PWR. Also, the various nuclide analysis technology would be identified and the technology which should be developed would be defined through this preliminary study.

Concerns with colloids, dispersed 1~1,000 nm particles, in the LILW repository are being raised due to their potential to enhance radionuclide release. Due to their large surface areas, radionuclides may sorb onto mobile colloids, and drift along with the colloidal transport, instead of being sorbed onto immobile surfaces. To prevent adverse implications on the safety of the repository, the colloidal impact must be evaluated. In this paper, colloid analysis done by SKB is studied, and factors to be considered for the safety assessment of colloids are analyzed. First, the colloid generation mechanism should be analyzed. In a cementitious repository, due to a highly alkaline environment, colloid formation from wastes may be promoted by the decomposition of organic materials, dissolution of inorganic materials, and corrosion of metals. Radiolysis is excluded when radionuclide inventory is moderate, as in the case of SKB. Second, colloid stability should be evaluated to determine whether colloids remain in dispersion. Stable colloids acquire electric charges, allowing particles to continuously repel one another to prevent coagulation. Thus, stability depends on the pH and ionic condition of the surroundings, and colloid composition. For instance, under a highly alkaline cementitious environment, colloids tend to be negatively charged, repelling each other, but Ca2+ ion from cement, acting as a coagulant, makes colloid unstable, promoting sedimentation. As in the case of SKB, the colloidal impact is assumed negligible in the silo, BMA, and BTF due to their extensive cement contents, but for BLA, with relatively less cement source, the colloidal impact is a potential concern. Third, colloid mobility should be assessed to appraise radionuclide release via colloid transport. The mobility depends on the density and size of colloids, and flow velocity to commence motion. As a part of the assessment, the filtration effect should also be included, which depends on pore size and structure. As in the case of SKB, due to static hydraulic conditions and engineering barriers, acting as efficient filters, colloidal transport is expected to be unlikely. In the domestic underground repository, the highly alkaline environment would lead to colloid formation, but due to high Ca2+ concentration and low flow velocity, colloids would achieve low stability and mobility, thus colloidal impact would be a minor concern. In the future, with further detailed analysis of each factor, waste composition, and disposal condition, reliable data for safety evaluation could be generated to be used as fundamental data for planning waste acceptance criteria.

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 prevent the release of radionuclides into the biosphere, disposal facilities for radioactive waste should be located to provide isolation from the accessible biosphere for tens of thousands to a million years after closure. During the period of interest, the constantly evolving natural environment and possible geological events of the site can cause disturbances to the containment function of the repository. Thus, for the long-term safety assessment of the repository, the possible long-term change of natural barrier should be considered. Due to the characteristics of radionuclides that transport mainly through the groundwater, understanding the long-term evolution of groundwater flow and geochemical properties is essential to assess the long-term changes in the natural barrier performance. The changes in characteristics of natural rocks and geological structures are one of the main factors that determine the hydrological and geochemical characteristics of the deep underground. In this study, we plan to develop a methodology to estimate these future geological evolutions in order to assess the possibility of hazardous events of the site that can affect hydrological or geochemical properties over the period of interest, and also in order to verify the change in the geological environment is within the safe performance range even after the period of interest. However, it is very unreliable to predict future changes in the natural environment because it is very heterogeneous, complex, and difficult to observe directly. For the preliminary study of the project, we reviewed cases of future evolution prediction researches with regard to the geological environment of disposal site and methods they applied to reduce the uncertainty of the prediction. The results will be used to establish basic data for future studies on the long-term evolution of hydraulic-mechanics performance of natural barrier and long-term evolution of geochemical performance around KURT site. In addition, it can contribute to construct long-term evolution scenario of the geological environment around future URL site.

A deep geological repository for disposal of high-level radioactive waste (HLW) consists of the canister, buffer material, and natural rock. If radionuclides leak from a disposal container, it can pass through buffer materials and rock, and move into the biosphere. Transport and migration of radionuclides in the rock differently were affected by the fracture type, filling minerals in the fracture, and the chemical and hydraulic properties of the groundwater. In this study, aperture distribution in fractured granite block was investigated by hydraulic test and CFD analysis. The fractured rock block (1 m × 0.6 m × 0.6 m), which is simulated as natural barrier, was prepared from Iksan, Jeollabuk-do. 9 test holes were drilled and packer system was installed to perform hydraulic test at the surface of fracture. 3D model simulated for aperture distribution of rock block was made using results of hydraulic test. And then, CFD analysis was performed to evaluate the co-relation between experiment results and analysis results using FLUENT code.

A radioactive waste repository consists of engineered barriers and natural barriers and must be safely managed after isolation. Geologic events in natural barriers should be categorized and evaluated according to their magnitude to assess the present and future stability of disposal. Among the longterm evolutionary elements of natural barriers, faults are a small portion of the Earth’s crust. Still, they play an important role in nuclide transport as conduits for fluids moving deep underground. In addition, the physical and chemical properties of fault rocks are useful for understanding the longterm and short-term behavior of faults. Paleomagnetic research has been used extensively and successfully for igneous, metamorphic, and sedimentary rocks. In addition, magnetic characterization of fault rocks can be used to describe faults or infer the timing of major geological events along fault zones. Components of magnetization defined in fault-breccias were attributed to chemical processes associated with hydrothermal mineralization that accompanied or post-dated tectonic activity along the fault. The study of magnetic minerals in fault rocks can be used as “strain indicators”, “geothermometers”, etc. This study is a preliminary test of magnetic properties using fault gouges. Fault gouges are not well preserved in typical terrestrial environments. Access to fresh gouges typically requires trenching through faults or sampling with a core drill. Fortunately, it is a magnetic property study using a fault gouge that exists on the inner wall of KURT (KAERI Underground Research Tunnel). This is to identify the motion history of the fault and, furthermore, to understand the stress structure at the time of fault creation. In addition, it can be presented as evidence for evaluating faults that may appear in future URL (Underground Research Laboratory).

High level radioactive waste disposal repository is faced thermos-hydro-mechanical-radioactive condition. Factors according to these complex conditions are measured using multiple sensors installed in the disposal repository to check integrity of the structure. Wires of the sensors can be potential pathways of groundwater and nuclide flow and these pathways accelerates bentonite saturation. Therefore, it is worth to developing wireless sensors buried in the bentonite buffer which can communicate without wires. In start of the study, widely-utilized wireless communication methods including WiFi and LoRa are tested using compacted bentonite blocks to estimate the performance of them. Compacted bentonite blocks are prepaired using di-press method with metal molds and the dry density of them are about 1.6 g/cm3. All wireless communication methods are well communicated through the bentonite blocks over 50 cm. The further experimental tests will be conducted with different dry density and water contents. The results of these experimental tests give a possibility of wireless communications in compacted bentonite buffer and will be utilized for the design of wireless sensor systems for the repository monitoring.

In-depth disposal of spent nuclear fuel means safe disposal of spent nuclear fuel by the concept of a multi-barrier system composed of an artificial barrier, an engineering barrier, and a natural barrier system of natural rock at a depth of less than 500 m underground. Disposal canisters are needed to store high-level waste in a deep environmental for a long time, and in order to demonstrate the performance of deep disposal canisters for spent nuclear fuel at underground research facilities (URL), it is intended to design disposal canisters and manufacture internal canisters. The internal canisters of spent nuclear fuel disposal canisters manufactured as a result of the study are combined with external copper canister technology and are directly used for demonstration of engineering barrier performance in underground facilities (URL) essential for final disposal of spent nuclear fuel. Disposal canister manufacturing technology and manufacturing process are used to manufacture disposal canisters for future final disposal projects in connection with domestic unique disposal systems. The quality inspection and quality management technology applied when manufacturing disposal canisters contribute to securing the soundness of disposal canisters that primarily maintain the safety of in-depth disposal by using them in the actual disposal business. By visually showing the development status of domestic disposal technology by displaying the prototype of disposal canisters manufactured as major achivements, the public can raise awareness of the domestic technology and safety of in-depth disposal of 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.

Backfill is one of the key elements of deep geological disposal. The backfill material is used to fill disposal tunnels and is mainly composed of swellable clay, preventing the migration of nuclide and structurally supporting the tunnel. The selection and application of backfill material are critical for the stable and efficient disposal of spent fuel. Therefore, it is essential to secure various candidate materials for backfill and to comprehensively understand the properties and behavior of these materials. Recently, the Korea Atomic Energy Research Institute has selected a candidate material called Bentonil-WRK and is evaluating its applicability. To utilize this material as backfill, the safety function of a mixed backfill concept, consisting of sand and Bentonil-WRK, was assessed. The swelling pressure was measured as a function of dry density for a bentonite/silica sand mix ratio of 3/7. The results showed that the swelling pressure ranged from 0.15 to 0.273 MPa, depending on the dry density, with higher dry densities resulting in higher swelling pressures. The measured swelling pressure met the target performance criteria suggested by SKB and Posiva (i. e., 0.1 MPa), but did not meet the design requirement for swelling pressure (i. e., 1 MPa). This indicate the need for further research after increasing the mass fraction of bentonite (e. g., mix ratio 4/6 or more). The results of this study are expected to be used in the selection of candidate backfill materials and the establishment of design guidelines for engineered barrier backfill.

A methodology is under development to reconstruct and predict the long-term evolution of the natural barrier comprising the site of radioactive waste disposal. The natural barrier must protect the human zone from radionuclides for a long time. So for this, we need to be able to restore the evolution of the bedrock constituting the natural barrier from the past to the present and to predict from the present to the future. A methodology is being studied using surface outcrop, tunnel face of KURT (KAERI Underground Research Tunnel), and drill core at KAERI (Korea Atomic Energy Research Institute). Among them, drill core is an essential material for identifying deep geological properties, which could not be confirmed near the surface when considering the geological condition of the repository in the deep part. In this study, we selected several qualitative and quantitative analyses to construct a deep lithological model from the disposal perspective. These were applied to drill core samples around the KURT. There are the dikes presumed the Cretaceous were intruded by Jurassic granitoids in the study area. Analyzing trace elements of each rock type in the study area classified through geochemical characteristics and microstructure in previous studies made it possible to obtain qualitative information on the petrogenetic process. In addition, synthesizing the quantitative numerical age allows for grasping the evolution of bedrock, including intrusion and cutting relationships. LAICPMS was used for determining the age of zircons in plutonic rocks. The highly reliable 40Ar-39Ar method was selected for volcanic rocks because it can correct the loss of Ar gas and obtain the values of two types of Ar isotopes in a single sample. As a result, it was possible to infer the formation environment of rocks through anomalies in specific trace element content. And according to the numerical ages, it was possible to support the known separated rock type found in previous studies or to present a quantitative precedence relation for unclassified rocks. These methods could be applied to reconstruct the long-term evolution of bedrock within natural barriers.

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.

Mixed-bed ion exchange resin consist of anion exchange resin and cation exchange resin is used to treat liquid radioactive waste in nuclear power plants. C-14 from heavy water reactors (HWR) is adsorbed on the anion exchange resin and is considered intermediate-level radioactive waste. The total amount of radioactivity of C-14 in spent ion exchange resin exceeds the activity limits for the disposal facility. Therefore, it is necessary to reduce the radioactivity through pre-treatment. There are thermal and non-thermal methods for the treatment of spent ion exchange resin. However, destructive methods have the problem of emitting off-gas containing radionuclides. To solve this challenge, various methods have been developed such as acid stripping, PLO process, activity stripping, thermal treatment and others. In this study, spent ion exchange resin (spent resin) was treated using microwave. The reaction characteristics of the resin to microwave were used to selectively remove the C-14 on the functional groups. Simulated spent anion exchange resin and spent resin from Wolseong NPP were treated with the microwave method, and the desorption rate was over 95%. An integrated process system of 1 kg/batch was built to produce operating data. After the operation of the process, characterization and evaluation of post-treatment for condensate water and adsorbent used in the process were performed. When the process system was applied to treat simulated spent resin and real spent resin, both showed a desorption rated of more than 97%. It means that the C-14 was successfully removed from the radioactive spent resin.