Domain decomposition method (DDM) has been widely employed for the numerical analysis of large-scale problems due to its applicability to parallel computing. DDM divides the modeling domain into a set of subdomains and obtains the entire solution iteratively until the values of each subdomain which are shared with other subdomains, such as boundary values, are converged. Therefore, in general, DDM is a memory-efficient iterative algorithm with inherent parallelism on the geometric level. APro, the process-based total system performance assessment model, aims for simulating the radionuclide transport considering coupled multi-physics phenomena occurring in large-scale geological disposal system, which are inevitably accompanied by huge memory burden. Therefore, DDM is applicable for the large-scale problem of APro and its performance in parallel computing needs to be examined. The DDM solvers provided by COMSOL which constitute APro can be classified into two methods. One is the overlapping Schwarz method that each subdomain overlaps its neighboring domains and the other is the Schur complement method that subdomains are non-overlapping and separated by boundary domains. For the Schwarz method, the additive, hybrid, multiplicative and symmetric methods can be selected according to the solution update scheme. And for the Schur method, the additive and multiplicative ordering options can be chosen for solving Schur complement system. In this study, the calculation efficiency of the DDM solvers in COMSOL and the applicability to the cluster environment were examined. In aspect of efficiency, the memory requirements with different number of subdomains and calculation schemes were compared in a single node. Then, the memory requirements with increasing number of disposal tunnels and deposition holes were investigated in multiple nodes. As a result, on the cluster environment, with the help of distributed memory architecture which enables efficient memory usage, the applicability of DDM solvers to the large-scale problem of APro was confirmed.

APro, developed by KAERI as a process-based total system performance assessment model, can simulate the radionuclide transport affected by thermal, hydraulic, mechanical and geochemical changes that may occurs in the engineering and natural barriers of a geological disposal system. APro targets a large-scale and heterogeneous 3D system that includes more than 10,000 boreholes located about 500 m underground and hundreds of fractures of different sizes distributed within an area of several km2. Simulating transport and reaction phenomena for such a system through the global implicit approach (GIA) may require considerable computational resources or be intractable in some cases. Therefore, APro adopts the sequential non-iterative approach (SNIA), one of the operator splitting (OS) methods, to separate the mass transport and reaction phenomena into independent problems. By using SNIA, the parallel computation performance in APro with multiple cores is expected to be improved. In this study, the effect of SNIA on the parallel computation performance was analyzed through a simple 1D reactive transport problem. Without SNIA, finite difference equations, discretized from the partial differential equations (PDEs) describing the reactive transport problem, have to be solved at once because all dependent variables are nonlinearly and spatially interconnected through reaction and mass transport terms. When the reaction and mass transport terms are separated through SNIA, the mass transport problem can be converted into independent linear equations for each chemical and the efficient linear system solver can be applied to each linear equation. In particular, since the reaction problem is changed to independent nonlinear equations for each node, the parallel computation performance can be greatly improved. To verify this, the 1D reactive transport problem was implemented in MATLAB, and SNIA and GIA were applied to solve the problem. As a result, there was no significant difference in results between SNIA and GIA for proper spatial and temporal discretization, which verified the accuracy of SNIA. In order to see the parallel computation performance, the calculation times for SNIA and GIA with increasing number of cores were measured and compared. As the number of cores increased, the SNIA calculation speed became faster than that of GIA, which verified that SNIA could improve parallel computation performance in APro. In the future, the effect of SNIA on the parallel computation performance will be verified for the numerical analysis of large-scale geological disposal systems.

The timescale for the post-closure safety assessment of a deep geological repository ranges from ten thousand to a million year. In such a long period of time, the biosphere inevitably undergoes changes. Therefore, the long-term evolution of a biosphere is recognized as an important issue in the post-closure safety assessment of a deep geological repository for spent fuels. In this study, we reviewed the approaches to address the long-term evolution of a biosphere. The major drivers of longterm evolution of a biosphere are the climate change and the resulting landscape development. They can affect the hydrogeological and hydrogeochemical characteristics of a biosphere, and then the radionuclide migration through the biosphere followed by the exposure doses for the critical groups. In addition, human activities and the social developments can affect the climate change resulting in the long-term evolution of a biosphere. To make a biosphere assessment, the long-term evolution scenarios for the biosphere should be formulated considering these climate change, landscape development, and human activities. In addition, features, events, and processes (FEPs) that affect the long-term evolution of a biosphere should be used. According to the Safety Case reports of Finland, the major long-term evolution scenario drivers of a biosphere are local sea-level change due to climate change and land use related to crop type, irrigation procedures, livestock, forest management, construction of a well, and demographics. The climate change causing the local sea-level change can be simulated using various earth system models such as CLIMBER-2, MPI/UW, and UVic and an icesheet model such as SICOPOLIS. The review results of this study and FEPs related to the climate change, the landscape development, and human activities will be used to formulate long-term evolution scenarios for the safety assessment of a deep geological repository for spent fuels.

The design of the high-level radioactive waste (HLW) repository is made for isolating the HLW from the groundwater system by using artificial and natural barriers. Granite is usually considered to be a great natural barrier for the HLW repository in various countries including Sweden, Canada, and Korea due to its low hydraulic permeability. However, many fractures that can act as conduits for groundwater and radionuclides exist in granite. Furthermore, the decay heat generated by the HLW can induce groundwater acceleration through the fracture. Since the direction, magnitude, and lasting time of the heat-induced groundwater flow can be differed depending on the fracture geometry, the effect of fracture geometry on the groundwater flow around the repository should be carefully analyzed. In this study, groundwater models were conducted with various fracture geometries to quantify the effect of various properties of fractures (or fracture networks) on the heat-induced groundwater flow. In all models, the pressure around the repository only lasted for a short period after it peaked at 0.1 years. In contrast, the temperature lasted for 10,000 years after the disposal inducing the convective groundwater flow. Single fracture models with different orientations were conducted to evaluate the variations in groundwater velocities around the repository depending on the fracture slope. According to the results, the groundwater velocity on the fracture was the fastest when the regional groundwater flow direction and the fracture direction coincided. In double fracture models, various inclined fractures were added to the horizontal fracture. Due to the intersecting, the groundwater flow velocity showed a discontinuous change at the intersecting point. Lastly, the discrete fracture network models were conducted with different fracture densities, length distributions, and orientations. According to the modeling results, the groundwater flow was significantly accelerated when the fracture network density increased, or the average fracture length increased. However, the effect of the fracture orientation was not significant compared to the other two network properties.

In recent years, the importance of the thermo-hydraulic-mechanical-chemical coupled processes is increasing in the performance assessment (PA) of the high-level radioactive waste repository. In the case of mechanical behavior, it is very important because it can affect fluid flow and radionuclide transport by changing the porosity and permeability of the medium. In particular, Excavation Damaged Zone (EDZ) should be considered essential in PA because the migration of radionuclide is affected by the enhanced hydraulic transmissivity and altered geomechanical behavior of EDZ. Furthermore, due to various thermo-hydraulic behaviors such as decay heat generated from radioactive waste, pore water pressure increase, and swelling pressure of bentonite buffer material, mechanical evolution is occurred which may change the size and physical properties of EDZ. Therefore, to solve this problem, analysis of coupled thermal-hydraulic-mechanical (THM) processes with the effect of long-term evolution of EDZ due to the mechanical behavior should be accompanied. In this study, numerical model for the long-term evolution due to mechanical behavior considering EDZ using the Adaptive Process-based total system performance analysis framework for a geological disposal system (APro) proposed by the Korea Atomic Energy Research Institute (KAERI). In the case of EDZ, the concept of Mazars’ damage evolution model was applied to simulate the behavior using the continuum model, and the change in hydraulic properties according to the degree of damage was considered. To investigate the importance of mechanical behavior in PA, the results were compared by performing numerical analysis according to the presence or absence of mechanical analysis. Finally, numerical analysis considering the mechanical evolution of EDZ was conducted using the model developed in this study to investigate the effect of EDZ.

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.

Multiple sorptive sites on natural illitic clays (e.g., frayed edge [FES], type II [TS], and planar sites [PS]) play an important role to diverse 137Cs immobilization in soil and aquifer environments. This study investigated the Cs sorption capabilities of 10 natural illitic clays at ranged Cs concentrations (1 ×10−7 to 1×10−3 mol·L−1) under various competing potassium concentration (distilled water to 1×10−1 mol·L−1). Additionally, multisite cation exchange model was performed to evaluate the best-fit sorption model and optimize the sorption capacities and affinities of multiple sorptive sites for Cs. Here, the experimental Cs sorption isotherms varied among 10 illtic clays, indicating different sorption capacities of Cs on illitic clays. The best-fit sorption model exhibited that variable Cs sorption of 10 illitic clays was significantly related to the sorption capacities at the FES (1.76 × 10−5 to 1.12×10−4 eq·kg−1), TS (1.59×10−3 to 9.76×10−3 eq·kg−1), and PS (2.14×10−2 to 1.51×10−1 eq·kg−1), respectively. The FES predominantly contributed to Cs sorption at low aqueous concentrations, whereas the TS and PS sorbed Cs at high concentrations. These sorption capabilities of multiple sorptive sites were correlated to illite contents and crystallinity of illitic clays, implicating that such parameters could be key factors to predict the Cs sorption for natural illitic clays in soil and aquifer environments. Finally, 1-D transport simulations represented that the severe Cs retardation occurred at low Cs concentration, implying that the FES predominantly affected to Cs transport in actual radioactive contamination sites (i.e., where low Cs concentration prevails), compared to the TS and/or PS.

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.

Comprehensive identification and systematic classification of all features, events and processes (FEP) that influence on the performance of a high-level radioactive waste disposal system is essential for safety assessment. Nuclear energy agency (NEA) has been developing and updating the standardized generic FEP list, so-called NEA international FEP list, which may be used as the basis to develop project-specific FEP lists to reflect diverse system and site characteristics in different countries. On the basis, Finland and Sweden have recently got licenses to construct spent nuclear fuel deep disposal facilities. Also in Korea, timely construction of a high-level radioactive waste disposal facility is an urgent issue for stable operation of nuclear power plants. For this end, a FEP list that properly considers for system and site characteristics of Korean high-level radioactive waste disposal facility needs to be developed. In this study, the most recent NEA international FEP list published in 2019 was comprehensively reviewed with focus on the structure of the classification system and the physicochemical mechanisms associated with the key elements. The obtained results will be used for the comparative analysis of domestic and oversea project-specific FEP lists and for the development of a generic FEP list relevant to Korean high-level radioactive waste disposal system.

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.

Barrier effect model developed by CRIEPI is used for the estimation of rate of radioactive material release from a transport cask submerged in the ocean. If the containment boundary of cask is broken in an accident during maritime transportation, the sea water comes into the cask cavity and the leaching of radioactive material occurs. If the release of radioactive material thorough the opening of the containment boundary of cask is less than the leaching rate of the radioactive material inside the cask, then the release rate is controlled by the saturation limit of the sea water inside the cask cavity. In this study, the release rate estimation using the barrier effect model is compared with the model used in other codes, such as MARINRAD. And by parameter study, important factors that affect the releaser rate are identified and prioritized. It is shown that the gap generated in the containment boundary is the key parameter that determine the release rate of the radioactive material and the leaching rate is the dominant parameter to determine the saturation time of the cavity sea water.

Currently, the development of evaluation technology for vibration and shock loads transmitted to spent nuclear fuel and structural integrity of spent nuclear fuel under normal conditions of transport is progressing in Korea by the present authors. Road transportation tests using surrogate spent nuclear fuel were performed in September, 2020 using a test model of KORAD-21 transportation cask and sea transportation tests were conducted from September 30 to October 4, 2021. In order to investigate amplification or attenuation characteristics, according to the load transfer path, a number of accelerometers were attached on a ship cargo hold, cradle, cask, canister, disk assembly, basket, and surrogate fuel assemblies and to investigate the durability of spent nuclear fuel rods, strain gages were attached on surrogate fuel assemblies. A ship named “JW STELLA” which has similar deadweight (5,000 ton) of existing spent nuclear fuel transportation ships was used for the sea transportation tests. The ship is propelled by 1,825 hp two main engines with two 4-bladed propellers. There are two major vibration sources in the ship. One is the vibration from waves and the other is the vibration from the engine and propeller system. The sensor locations on the ship were determined considering the vibration sources. The sea transportation test was performed for 5 days, the test data were measured successfully. The ship with the test model was departed from Changwon and sailed to Uljin, sailed west to Yeonggwang and then returned to Changwon. In addition to sailing on a designated test route, circulation test, braking/acceleration test, depth of water test, and rolling test were conducted. As a result of the preliminary data analysis of the sea test, power spectral densities and shock response spectrums were obtained according to the different test conditions. The vibratory loads caused by the wave mainly occurred in the frequency range of 0.1 to 0.3 Hz. The vibratory loads caused by the propeller occurred near the n/rev rotating frequencies, such as 5, 10, 20 Hz etc. However, those frequencies are far from the natural frequencies of local mode of the fuel rods, so it is considered that the vibratory loads from the wave and the propeller do not have a significant influence on the structural integrity of the fuel rods. Among all the test cases, maximum strain occurred at SG31 near the bottom nozzle on the test; the magnitude was 73.62 micro strain. Based on the analyzed road and sea transportation test data, a few input spectra for the shaker table test will be obtained and the shaker table test will be conducted in 2022. It is expected that the detailed vibration characteristics of the assembly which were difficult to identify from the test results can be investigated.

The temperature of the spent fuel cladding is the basis for the evaluation of integrity. It is almost impossible to directly measure the temperature of spent nuclear fuel. Because spent nuclear fuel is dangerous. We are preparing a test to measure the cladding temperature with an equivalent fuel assembly by simulating the characteristics of spent nuclear fuel. PLUS7 was selected as the test target in consideration of the amount of generation, thermal water retention, residual moisture content, and manufacturability of domestic spent nuclear fuel. The nuclear fuel assembly is planned to be manufactured in two main ways. Except for the cladding tube that simulates decay heat, the structure will be manufactured by KEPCO Nuclear Fuel, and fuel rods and canisters will be manufactured by SUKEGAWA Electric in Japan. The same nuclear fuel assemblies as commercial skeleton will be applied. The temperature of the fuel cladding will be measured by attaching a thermocouple directly to the surface of the cladding tube. The canister is composed of a basket, a basket supporter, a heater and drain tube, a lead, and an observation window. The working fluid is water and helium, and the maximum pressure inside the canister is 1.1 MPa and the minimum pressure is 0.05 kPa. The maximum temperature of the surface of the cladding was designed to be 500°C, and the maximum temperature of the sealing to keep airtightness was designed to be 250°C. To satisfy this condition, we plan to evaluate the leak rate below 10−5 std.cm3·s−1, which is equivalent to helium tightness. The maximum heat of decay per fuel rod is 13 W, and one assembly is up to 3 kW. Production of the test equipment is expected to be completed in the first half of next year, and testing is scheduled to begin in the second half of next year. The test will evaluate all environments that the spent nuclear fuel may experience, such as dry normal conditions, abnormal conditions, wet conditions, and dry conditions. All data will be used for interpretation verification purposes.

With the enhancement of the spatial resolution of satellite imagery (1 m or less), the satellite image analysis has been considered as the indispensable means for remote sensing of nuclear proliferation activities in the restricted access areas such as North Korea. Notably, in the case of an open-pit uranium mine, e.g. the Pyongsan uranium mine, the mining activity can be presumed if detecting the location and extent uranium tailing piles near shafts within temporal images. Several studies have researched on the target detection for minerals of interest such as limestone and coal to evaluate the economic activities by utilizing similarity measures, e.g., a spectral angle mapper and a spectral information divergence (SID). Thus, this paper presented a systematic change detection methodology for monitoring the uranium mining activity in the Pyongsan uranium mine with a similarity measure of SID. The proposed methodology using the target detection results consists of the following five steps. The first step is to acquire stereo images of areas of interest for change detection. The second step is to preprocess the stereo images as following measures: (i) the QUick Atmospheric Correction and the image-to-image registration with ENVI and (ii) the Gram-Schmidt pansharpening. The third step is to extract spectral information for minerals of interest, i.e., uranium tailing piles, by sampling pixels within the reference image. It is based on the satellite analysis report for the Pyongsan uranium mine by CSIS, which specified the location of the uranium tailing piles. As the fourth step, the target detection for uranium tailing piles was performed through the similarity measure of SID between the extracted spectral information and the spectral reflectance of the image. In the fifth step, the change detection was processed using the multivariate alteration detection algorithm, which compares the target detection results by canonical correlation analysis. Furthermore, this paper evaluated the performance of the proposed methodology with the change detection accuracy assessment index, i.e., the area under a receiver operating characteristic curve. In conclusion, this paper suggests the systematic change detection methodology utilizing time series analysis of target detection for uranium tailing piles, which can save time and cost for humans to interpret large amounts of satellite information at the restricted access areas. As future works, the feasibility of the proposed methodology would be investigated by analyzing distribution of minerals of interest regarding nuclear proliferation at Yongbyon, which has the historical events of suspicious nuclear activities.

In the previous study, the types of North Korea’s strategic items, foreign trading partners, and export items were investigated. From North Korea’s typical trade paths, it is possible to predict the paths through which North Korea’s strategic items are illegally exported upon denuclearization. Trading partners of North Korea are the potential importing countries or end-users of strategic items, which can be disguised or concealed as if it is general export items during typical export procedures. So, in this study, transfer paths of North Korea’s export items are examined by utilizing KOTRA statistics, including item type HS code and its total price. Also, AnyLogic, a comprehensive simulation modeling tool, the simulation will be conducted to identify the paths for illegal transfer and calculate the time required. The information on North Korea’s trading partners and items is used for establishing export scenarios in which strategic items are transferred to other countries through North Korea’s ports, airports, railroads, and roads. To be specific, China, Russia, and South Korea, countries that share a border with North Korea, export items transported only by land; the items will arrive first in the referred three countries. Since the types of items, North Korea transacts with each country are different, the total amount and frequency of transactions are different; the probability of strategic items being included in general export items and transferred during customs clearance also varies. Even if it does not border North Korea, North Korea can export items through ports to countries adjacent to the coastline, and North Korea can even export items to any country by airspace even if it is not adjacent to the coastline. So, all publicly open ports, airports, railways, and roads are surveyed. Their geographic information, such as EPSG 4326 and EPSG 3857 coordinate system, are applied to confirm and visualize valid export paths starting from North Korea. In conclusion, effective export paths in North Korea are identified based on North Korea’s each major transportation hub by using AnyLogic simulation. It is possible to predict the paths through which North Korea’s strategic items will be transferred by combining information on major export items and countries that North Korea mainly transacts with.

Investigating major trading partners and items with North Korea is informative in terms that it can predict the path through which North Korea’s strategic items will transfer to non-nuclear-weapon states when North Korea denuclearizes. By analyzing North Korea’s trading partners and the items, it is possible to identify the relevant countries through which items arrive from the first importing country to the end-user in the process of exporting items and to predict the way how North Korea disguise or conceal their strategic items among general items during normal export procedures. As of 2020, North Korea’s major trading partners are China, Russia, Vietnam, India, Nigeria, and Switzerland. Compared to 2019, Mozambique, Tanzania, Ghana, and Thailand entered the top 10, while Brazil, Bangladesh, Pakistan, and South Africa pushed out of the top 10. North Korea’s trade dependence on China accounts for 88.2%, making it the largest trading partner for years, and it shows that North Korea is mainly conducting trade with Asian and African countries. North Korea’s most important export items are mineral products (HS 25-27) and steel & metal products (HS 72-83) and the most significant import items are mineral products (HS 25-27) and oils & fats & prepared foods (HS 15-24). In 2017, due to UN Security Council sanctions for North Korea’s international ballistic missile (ICBM) test-fire, North Korea’s exports from 3 billion dollars fell by 90% to less than 300 million dollars. This is the result of most of North Korea’s major export items included in the export ban, and changes have occurred in its export items. In 2020, export fell to less than 100 million dollars due to border lockdown measures to prevent the spread of COVID-19, which also affected the change of North Korea’s major export items. Although North Korea does not officially publish its foreign trade statistics, in order to review North Korea’s trade information, KOTRA statistics are utilized. KOTRA statistics provide only two digits of HS code number, so it is challenging to identify detailed item classification. Moreover, these statistics are based on the export amount, so it is difficult to determine the exact quantity of export items. It is expected that information on North Korean trading partners and items will be used to predict potential transferable export methods of North Korea’s strategic items when North Korea denuclearizes.