Source localization technique using acoustic emission (AE) has been widely used to track the accurate location of the damaged structure. The principle of localization is based on signal velocity and the time difference of arrival (TDOF) obtained from different signals for the specific source. However, signal velocity changes depending on the frequency domain of signals. In addition, the TDOF is dependent on the signal threshold which affects the prediction accuracy. In this study, a convolutional neural network (CNN)-based approach is used to overcome the existing problem. The concrete block corresponding to 1.3×1.3×1.3 m size is prepared according to the mixing ratio of Wolseong low-to-intermediate level radioactive waste disposal concrete materials. The source is excited using an impact hammer, and signals were acquired through eight AE sensors attached to the concrete block and a multi-channel AE measurement system. The different signals for a specific source are time-synchronized to obtain TDOF information and are transformed into a time-frequency domain using continuous wavelet transform (CWT) for consideration of various frequencies. The developed CNN model is compared with the conventional TDOF-based method using the testing dataset. The result suggests that the CNN-based method can contribute to the improvement of localization performance.

The radwaste facility management team is preparing for clearance of 4 MCAs in The Radwaste Form Test Facility (RFTF). The targeted waste was used for clearance level radioactive waste sample analysis and has been used for this purpose since the early 2000s. Due to the characteristics of clearance level radioactive waste, the concentration of radioactivity is very low and MCA is used with Marinelli beakers the possibility of contamination is low. Moreover, the radiation detector should not be contaminated with radioactive materials, it should be less than the clearance level. These detectors were considered surface contamination materials. To detect the contaminated spot of each material, we scanned the whole surface of a material with a gamma survey meter. After that, we took a sample from 1×1 m2 and a total of 30 samples from each MCA. The wiped filter paper was analyzed with alpha, beta low background counting systems. The results of the analysis of the smear sample of total alpha and beta nuclide radioactivity were less than MDA (α: 2.88×10−5 Bq·cm−2, β: 3.07×10−5 Bq·cm−2). The major nuclide in this facility is Co-60 and Cs-137 therefore we analyzed gamma nuclide activity with HPGe. The maximum specific activity was Co-60: 2.31×10−5 Bq·cm−2, Cs-137: 1.96×10−6 Bq·cm−2. If it is satisfied with the clearance criteria, detectors will be reused at the Radioactive Waste Treatment Facility (RWTF) room # 7251 uncontrolled area.

Low and intermediate radioactive wastes in South Korea have been disposed in Wolsong Low and Intermediate Level Radioactive Waste Disposal Center (WLDC), Gyeongju. This repository structure is planned to be operated few hundred years while toxicity of the waste is sufficiently decayed. The structural integrity of the repository is required to protect the waste in safe. The integrity of the structure is commonly estimated using acoustic emission (AE) method. The integrity of the structure using AE is obtained by following process: 1) Estimation of maximum acoustic crack energy of the structure, 2) Acoustic signal measurement and filtering, and 3) Measurement of simultaneous acoustic cracking energy. The damage of the structure can be obtained from cumulative cracking energy from the structure divided by the predicted maximum cracking energy of the structure. Estimation of maximum cracking energy is gained by the specimens whose components are identical to the repository structure. The cracking energy of the different specimens are obtained during uniaxial compressive test and volume of the specimen is calculated. Then, the fractal coefficient for the structure is obtained and the maximum crack energy of the target structure can be calculated. The specimens whose diameters vary from 50 mm to 150 mm and heights are twice of the diameter are made with same recipe of WLDC silo concrete. The uniaxial compression test is conducted with loading rate of 0.1 mm·min−1. The fractal coefficient is obtained by least square method from the volume-cumulative energy relationship.

Low to intermediate radioactive waste disposal concrete structures are subjected to coupled hydromechanical conditions and the identification of structural damage is crucial to ensure safe long-term disposal. Different damage models for concrete and the surrounding rock can affect the damage characteristics of radioactive waste disposal structures. In this study, the effects of different rock damage models are applied to the hydro-mechanical-damage coupled structural analysis of the Wolseong Low and Intermediate Level Radioactive Waste Disposal Center silo. A two-dimensional model of the disposal silo was modeled using the finite element analysis software COMSOL and the Mazars’ damage model was applied to the silo concrete. The Mazars’ model parameters were obtained from uniaxial compression and tensile tests on cylindrical concrete specimens after 28 days of water curing and further 32 days of wet curing at 75°C). The COMSOL embedded Richards equation module was used to simulate hydraulic analysis. Structural loading due to waste disposal was applied at the bottom of the silo structure and the damage evolution characteristics were investigated. The non-linear mechanical rock behavior obtained from laboratory tests (Hoek-Brown criterion, resonant column test, Mazar’s damage model) and field tests (Goodman Jack) were input to assess the effects of different rock damage models. The results highlight the importance of structural damage consideration when assessing the long-term stability and safety of underground radioactive waste disposal structures under coupled hydro-mechanical conditions.

According to the Atomic Energy Act of Korea, radioactive waste can be cleared when it meets the criteria, less than 10 uSv·y−1 for individual dose and 1 person · Sv·y−1 for collective dose. Consequently, it is necessary to evaluate radiation dose to get permission for regulatory clearance from the regulatory body of Korea. Several computational programs can be used for dose calculation depending on disposal methods such as landfill, incineration, and recycling. As for incineration, the effects of radionuclide emitted during combusting radwaste have to be considered to figure out exposure dose. In this study, GASPAR code is described to assess exposure dose from effluents released to the atmosphere during incinerating combustible radioactive wastes for regulatory clearance. GASPAR is the code programmed by Radiation Safety Information Computational Center at Oak Ridge National Laboratory for computing annual dose due to radioactive effluents released from a nuclear power plant to the atmosphere during routine operation. The calculating methods of the code are based on the mathematical model of U.S. NRC regulatory guide 1.109, about beta and gamma radiation from noble gas in semi-infinite plume, radioiodine, and particulates. GASPAR evaluates both individual dose and population dose. The considering pathways are composed of external exposure by plume and ground deposition of effluents, and internal exposure as a result of inhalation and food ingestion. Since the calculation model of GASPAR requires various variables about the radionuclide and disposal site, the accuracy of the results is decided by inputted values. The program contains the default values to parameters such as the humidity, fraction of deposition, and storage time of foods. However, to get permission, it is important to use the appropriate data representing the condition of the combustion scenario as substitutes for the default since the values are localized to the country where the code was developed. Therefore, dose assessment by GASPAR code can be applied for regulatory clearance by incineration, when reliable values depending on the disposal plan inputted.

Expansive clays (for examples, bentonites) are favored as buffer and backfill materials because of their low hydraulic conductivity, high swelling potential, and good mechanical properties, and are installed in highly compacted blocks in repositories. Compacted expansive clays have a dual-structure system: macrostructural system which is a complex of clay aggregates with the inter-aggregate pores (macropores) which can be filled by either liquids or gases; microstructural system with the intraaggregate pores between or within clay particles (micropores) which is usually considered to be saturated by liquid. Understanding the dual-strucure system of expansive clays is essential for characterizing and modeling multiphysics (stress-strain, swelling pressure, etc.) in buffers and backfills. Existing multiphysics studies of expansive clays, as in non-expansive soils, were mostly conducted with a single structure approach based on the behavior of macropores, and there have been limitations in the comprehensive interpretation and modeling of experimental results. However, with the recent development of measurement techniques, a lot of available information on the pore structure of compacted expansive clays has been reported, and with the results, a dual-structure approach considering both microstructural and macrostructural systems has been increasingly applied to improve the modeling of multiphysics of expansive clays. This study reviewed the dual-structure system of compacted expansive clays, analyzed previous studies on its evolution according to hydromechanical loading (loading-unloading and wetting-drying paths), and based on these, intended to provide technical knowledge and information needed for multiphysics research of expansive clays-based buffer and backfill for the KRS repository.

Gases such as hydrogen can generate from the disposal canister in high-level radioactive waste disposal systems owing to the corrosion of cooper container in anoxic conditions. The gas can be accumulated in the voids of bentonite buffer around the disposal canister if gas generation rates become larger than the gas diffusion rate of bentonite buffer with the low-permeability. Continuous gas accumulations result in the increase in gas pressure, causing sudden dilation flow of gases with the gas pressure exceeding the gas breakthrough pressure. Given that the gas dilation flow can cause radionuclide leakage out of the engineered barrier system, it is necessary to consider possible damages affected by the radionuclide leakage and to properly understand the complicated behaviors of gas flow in the bentonite buffer with low permeability. In this study, the coupled hydro-mechanical model combined with the damage model that considers two-phase fluid flow and changes in hydraulic properties affected by mechanical deformations is applied to numerical simulations of 1-D gas injection test on saturated bentonite samples (refer to DECOVALEX-2019 Task A Stage 1A). To simulate the mechanical behavior of microcracks which occur due to the dilation flow caused by increase in gas pressure, a concept of elastic damage constitutive law is considered in the coupled hydro-mechanical model. When the TOUGH-FLAC coupling-based model proposed in this study is applied, changes in hydraulic properties affected by mechanical deformations combined with the mechanical damage are appropriately considered, and changes in gas injection pressure, pore pressures at radial filters and outlet, and stress recorded during the gas injection test are accurately simulated.

High level nuclear waste (HLW) is surely disposed in repository in safe by being separated from human life zone. Deep geological disposal method is one of the most potent disposal method. Deep geological repository is exposed to high pressure and groundwater saturation due to its depth over 500 m. And it is also exposed to high temperature and radiation by spent fuels. Thus, HLW repository suffers extremely complex thermo-hydro-mechanical-radioactive condition. Long-term integrity of repository should be verified because the expected lifetime of the repository is over 10,000 years. However, the integrity of monitoring sensors are not reach the endurance lifetime of the repository with present technology. And the disposal condition, thermo-hydro-mechanical-radioactive, should shorten the estimated lifetime of the monitoring sensors. Therefore, it is necessary to improve the long-term integrity of the monitoring sensors. Although long-term tests are required to identify the prolonged durability of monitoring sensors, accelerated tests can help curtail test period. Accelerated tests is classified into accelerated stress test and accelerated degradation test and their methodology and theories are investigated. Their tests are design and proceed by following process: 1) identify failure modes, 2) select accelerated stress parameter, 3) Determine stress level, 4) Determine testing time and number of specimens, 5) Define measurement paremeter and failure criteria, 6) Suggest measurement method and measurement duration. Literature reviews were conducted to identify the influence of the disposal conditions such as thermo-hydro-mechnical-radioactive on integrity of material and monitoring sensors. The investigated data reported in this paper will be utilized to verify the improvement of integrity of monitoring sensors.

Discontinuum-based numerical methods can contain the multiple discontinuities in a model and reflect the thermal, hydraulic and mechanical characteristics of discontinuities. Therefore, discontinuum methods can be appropriate to simulate the model which require the detailed analysis of the coupled thermo-hydro-mechanical processes in fractured rock such as geothermal energy, CO2 geo-sequestration, and geological repository of the high-level radioactive waste. TOUGH-3DEC, the three-dimensional discontinuum simulators for the coupled thermo-hydro-mechanical analysis, was developed by linking the integral finite difference method TOUGH2 and the explicit distinct element method 3DEC to describe the coupled thermo-hydro-mechanical processes in both porous media and discontinuity. TOUGH2 handles thermo-hydraulic analysis by the internal simulation module, and 3DEC performs mechanical study based on the constitutive models of porous media and discontinuity with coupling the thermal and hydraulic response from TOUGH2. The thermal and hydraulic couplings are the key processes and should be carefully verified by sufficient cases, so this study performed the thermomechanical and hydro-mechanical simulations which are modelling the analytic solutions including the uniaxial consolidation, fracture static opening, and the heating of a hollow cylinder problems. Each thermo-mechanical and hydro-mechanical verification case is also validated by comparing with the results of the other continuum and discontinuum-based numerical methods. TOUGH-3DEC results follow the analytic solutions and show better accuracy than the continuum-based numerical methods in the static fracture opening problem. The developed TOUGH-3DEC simulator can be expanded to coupled thermo-hydro-mechanical-chemical analysis in fractured rock mass, and the simulator needs to be verified by more complicated coupled processes problems which require in the chemical coupling.

The backfill close the deep geological disposal system by filling the disposal tunnel and the connecting tunnel after the installation of buffer in the disposal hole. SKB and Posiva have established and designed the safety function of the backfill for the common goal of the deep geological disposal system. The safety function of backfill material has been set hydraulic conductivity of less than 10−10 m·s−1, a swelling pressure of 0.2 MPa, a compressive modulus of 10 MPa or a buffer density of 1,950 kg·m−3 or more, and freezing resistance. For the selection of the optimum backfill material, SKB and Posiva developed the concept of the backfill and evaluated the candidate that satisfies the requirements in four steps. In the first step, the performance and function that the backfill material should have were conceptualized. For the second step, laboratory tests and in-depth analysis of the candidate material properties were conducted. At this step, the focus has been on testing with the concept of the block method, using key candidate materials. In step 3, laboratory and large-scale experiments were performed to test engineering feasibility. In addition, design specifications for backfill materials were set based on site conditions, installation methods, and short- and long-term functions of materials. In Korea, it is only now in the step of selecting the concepts of the safety function. Therefore, it is necessary to benchmark the development process based on the previous studies of SKB and Posiva. In this study, candidate materials, experimental methods, and results were analyzed. As a result, the research steps and conditions for the selection of the optimum backfill material were reviewed. Using this study, the research steps of domestic backfill was suggested to develop within a short time for the Korean deep geological disposal system.

Corrosion cells that simulates engineering barrier system have been stored in an aerobic KURT environment for 10 years, which were recovered and dismantled in 2021. The test specimens were compressed copper (Com. Cu), Cold spray copper (CSC Cu), Ti Gr.2, STS 304, and Cast nodular iron. The specimens were buffered by compact Ca-type Gyeongju bentonite (KJ-I) and compact Na-type Wyoming bentonite. And the corrosion cells were exposed to KURT groundwater at 30°C for about 10 years (3,675 days). As a result of the long-term experiment in aerobic environment, it was confirmed that Na-bentonite is more advantageous for inhibiting corrosion than Ca-bentonite. The corrosion thickness of the most specimens in Ca bentonite was slightly lower than in Na bentonite until the initial 500 days, but after 10 years, the corrosion thickness of copper and cast iron specimens in Na bentonite was clearly lower. The corrosion thickness of the copper specimen in Na bentonite was very low about 0.5 um in both Com. Cu and CSC Cu. Moreover, the corrosion thickness in Ca bentonite was very high about 4 um for Com. Cu and 6 um for CSC Cu. In the case of cast iron, the corrosion thickness in Na bentonite was about 13 um, and 15 um in Ca bentonite. The common feature of copper and cast iron specimens in Ca bentonite, which showed a high corrosion thickness, is the forming of a white mineral deposition layer on the specimen surface, which was presumed to be some kind of feldspar. On the other hand, it was found that the STS304 and Ti specimens were hardly corroded even after 10 years. In conclusion, when a white mineral deposition layer was formed on the specimen surface, the corrosion thickness always increased sharply than before, and thus it was estimated that the generation of the mineral deposition layer cause the increase of bentonite permeability, and rather the weakening of existing passive corrosion film.

A geological repository system consists of a disposal canister with packed spent fuel, buffer material, backfill material, and intact rock. Among these, the bentonite buffer is one of the most important components to assure the safe disposal of high-level radioactive waste (HLW). As the bentonite buffer is installed as a block type, it is important to fabricate homogeneously. Generally, floating die method and cold isostatic press (CIP) method are used to fabricate bentonite blocks. In this paper, two bentonite blocks were produced using float die method at first, and CIP method was additionally applied to just one block. After that, several samples were cored from two blocks. The dry density and water content of several samples produced from two blocks were measured.

With the increase of temporarily-stored spent radioactive fuels, there is an increasing necessity for the safe disposal of high-level radioactive waste (HLW). Among various methods for the disposal of HLW, a deep geological disposal system is adapted as a HLW disposal strategy in many countries. Before the construction of a repository in deep geological condition, a performance assessment, which means the use of numerical models to simulate the long-term behavior of a multi-barrier system in HLW repository, has been widely performed to ensure the isolation of radionuclides from human and related environments for more than a million years. Meanwhile, Korea Atomic Energy Research Institute (KAERI) is developing a process-based total system performance assessment framework for a geological disposal system (APro). To improve the reliability of APro, KAERI is participating in DECOVALEX-2023 Task F, which is the international joint program for the comparison of the models and methods used in deep geological performance assessment. As a final goal of Task F, the reference case for a generic repository in fractured crystalline rock is described. The three-dimensional generic repository is located in a domain of 5 km in length, 2 km in width, and 1 km in depth, and contains an engineering barrier system with 2,500 deposition holes in fractured crystalline rock. In this study, a numerical simulation of the reference case is performed with COMSOL Multiphysics as a part of Task F. The fractured crystalline rock is described with the discrete fracture matrix (DFM) model, which expresses major deterministic fractures explicitly in the domain and minor stochastic fractures implicitly with upscaled quantities. As an output of the numerical simulation, fluid flow at steady-state and radionuclide transport are evaluated for ~106 years. The result shows that fractures dominate the transport of radionuclides due to much higher hydraulic properties than rock matrix. The numerical modeling approaches used in this study are expected to provide a basis for performance assessment of nuclear waste disposal repository located in fractured crystalline rock.

Geologic disposal of high-level radioactive waste is considered the most effective method to isolate high-level radioactive waste from the biosphere. A high-level radioactive waste repository is designed to be placed at a deep depth and generally consists of canisters, buffer material, and host rock. In the disposal system, the heat from the canister occurs for millions of years due to the long half-life of the high-level radioactive waste, and the heat induces vaporization of groundwater in the buffer material. The resaturation process also occurs due to groundwater inflow from the host rock by the hydraulic head and capillarity. The saturation variation leads to the heat transfer and multi-phase flow in the buffer material, and thermal pressurization of groundwater due to the heat affects the effective stress change in the host rock. The stress change can make the porosity and permeability change in the flow system of the host rock, and the flow system affects the nuclide migration to the biosphere. Therefore, it is crucial to understand the complex thermo-hydro-mechanical-chemical (THMC) coupled behavior to secure the repository’s long-term safety. DECOVALEX is an international cooperating project to develop numerical methods and models for predicting the THMC interactions in the disposal systems through validation and comparison with test results. In Task C of DECOVALEX-2023, nine participating groups (BGR, BGE, CAS, ENSI, GRS, KAERI, LBNL, NWMO, Sandia) models the full-scale emplacement (FE) experiments at the Mont Terri underground rock laboratory and focus on understanding pore pressure development, heat transfer, thermal pressurization, vaporization and resaturation process in the disposal system. In the FE experiment, three heaters generated heat with constant power for five years at a 1:1 scale in the emplacement tunnel based on Nagra’s reference repository design. KAERI used OGS-FLAC3D for the numerical simulation, combining OpenGeoSys for TH simulation and FLAC3D for M simulation. We generated a full-scale three-dimensional numerical model with a dimension of 100 by 100 by 60 meters. The pressure and temperature distribution were well simulated with the host rock's anisotropy. Based on the capillarity, we observed vaporization and resaturation in the bentonite under the twophase flow system. We plan to compare the simulation results with the field data and investigate the effect of input parameters, including thermal conductivity and pore compressibility affecting the thermal and flow system.

Deep geologic disposal of high-level nuclear wastes (HLW) requires intensive monitoring instrumentations to ensure long-term security. Acoustic emission (AE) method is considered as an effective method to monitor the mechanical degradation of natural rock and man-made concrete structures. The objectives of this study are (a) to identify the AE characteristics emitted from concretes as concrete materials under different types of loading, (b) to suggest AE parametric criteria to determine loading types and estimate the failure stage, and finally (c) to examine the feasibility of using AE method for real-time monitoring of geologic disposal system of HLW. This study performs a series of the mechanical experiments on concrete samples simultaneously with AE monitoring, including the uniaxial compression test (UCT), Brazilian tensile test (BTT) and punch through shear test (PTST). These mechanical tests are chosen to explore the effect of loading types on the resulting AE characteristics. This study selects important AE parameters which includes the AE count, average frequency (AF) and RA value in the time domain, and the peak frequency (PF) and centroid frequency in the frequency domain. The result reveals that the cumulative AE counts, the maximum RA value and the moving average PF show their potentials as indicators to damage progress for a certain loading type. The observed trends in the cumulative AE counts and the maximum RA value show three unique stages with an increase in applied stress: the steady state stage (or crack initiation stage; < 70% of yield stress), the transition stage (or damage progression stage; 70–90% of yield stress) and the rising stage (or failure stage; > 90% of yield stress). In addition, the moving average PF of PTST in the early damage stage appears to be particularly lower than that of UCT and BTT. The loading in BTT renders distinctive responses in the slope of the maximum RA–cumulative AE count (or tan ). The slope value shows less than 0.25 when the stress is close to 30% of BTT, 60% of UCT and 75% of PTST and mostly after 90% of yield stress, the slope mostly decreases than 0.25 in all tests. This study advances our understanding on AE responses of concrete materials with well-controlled laboratoryscale experimental AE data, and provides insights into further development of AE-base real-time diagnostic monitoring of structures made of rocks and concretes.

A methodology is under development to restore and predict the long-term evolution of the natural barrier comprise the site of radioactive waste disposal for surface geological outcrop, tunnel face and drill core. Considering the condition that the radioactive waste repository should be located in the deep part, the drill core is an important subject that can identify deep geological properties that could not be confirmed near the surface. In this study, we investigated proper age dating methods to construct lithological model of the disposal site with regard to the long-term safety. Also, preliminary age dating locations were selected using the lithological distribution results by depth through geochemical and micro-structural analysis for the deep drill cores excavated around KURT. In the study area, the dikes presumed the Cretaceous were intruded by Jurassic granites. As for the granotoids, U-Pb age dating for zircon, which is resistant to deformation or metamorphism and has loss, is often used. In the case of the dikes, K-Ar and 40Ar/39Ar age dating for the argon captured in the rocks after magmatism is often used. Through U-Pb zircon ages of KURT site granotoids, we expect to solve the clustering problem (granite and granodiorite), which is different from precious chemical analysis (XRF) results and TAS-diagrams. 40Ar/39Ar age dating to be used for the dikes is suitable for the perspective of lithological model of the disposal site. Because, it can compensate for accuracy problems such as sample heterogeneity in K-Ar age dating and is used for volcanic rocks. In the further study, we plan to determine the appropriate sampling locations by the selected age dating methods from the perspective of disposal in this study.

Deep geological disposal with multiple barriers composed of engineered and natural barriers has been considered as the most suitable disposal method for high level nuclear wastes. In terms of the geological evaluation factors, brittle structures such as fractures and faults should be characterized around the repository site, because radionuclides transfer mainly with groundwater in the subsurface and groundwater flows through discontinuous brittle structures. The geological survey for the characterization of deep geological repository sites is widely conducted by narrowing the survey area from regional scale down to local scale, which could be divided into three steps: 1) using remote sense or geophysical survey, 2) trench and drill core logging including field survey based on the first step, 3) detailed geological survey in the tunnel. In this study, we analyzed the distribution of geological structures to derive the history of brittle deformation in and around the KURT (KAERI Underground Research Tunnel) site located in the KAERI (Korea Atomic Energy Research Institute). The bedrock of the KURT site is mainly consist of Jurassic two-mica granite, which is extensively intruded by andesitic dikes of Cretaceous with N-S to NE-SW strikes. The two-mica granite in the study area was deformed in a ductile deformation environment and has been overprinted by major geological structures such as faults, dikes, veins, and joints. From this study, we identified 8 brittle deformation events based on the cross-cutting relationship among the geological structures, which are obtained from the analyses in and around the KURT. In order to evaluate the reactivation and fluid flow potential of brittle structures, it is essential to determine the characteristics and ages of the brittle structures and the composed rocks around the site.