For countering nuclear proliferation, satellite imagery is being used to monitor suspicious nuclear activities in inaccessible countries or regions. Monitoring such activities involves detecting changes over time in nuclear facilities and their surroundings, and interpreting them based on prior knowledge in terms of nuclear proliferation or weaponisation. Therefore, analysts need to acquire and analyze satellite images periodically and have an understanding of nuclear fuel cycle as well as expertise in remote sensing. Meanwhile, as accessibility of satellite information has been increasing and accordingly a large amount of high-resolution satellite images is available, a lack of experts with expertise in both fields to perform satellite imagery analysis is being concerned. In this regard, the Institute of Korea Nonproliferation and Control (KINAC) has developed a prototype of semi-automatic satellite imagery analysis system that can support monitoring of potential nuclear activities to overcome the limitations of professionals and increase analysis efficiency. The system provides a satellite imagery database that can manage acquired images, and the users can load images from the database and analyze them in stages. The system includes a preprocessing module capable of resizing, correcting and matching images, a change detection module equipped with a pixel-object-based change detection algorithm for multi-temporal images, and a module that automatically generates reports with relevant information. In particular, this system continuously updates open-source information database related to potential nuclear activities and provides users with an integrated analytics platform that can support their interpretation by linking related images and textual information together. As such, the system could save time and cost in processing and interpreting satellite images by providing semi-automated analytic workflows for monitoring potential nuclear activities.

Nuclear power plant decommissioning generates significant concrete waste, which is slightly contaminated, and expected to be classified as clearance concrete waste. Clearance concrete waste is generally crushed into rubble at the site or a satellite treatment facility for practical disposal purposes. During the process, workers are exposed to radiation from the nuclides in concrete waste. The treatment processes consist of concrete cutting/crushing, transportation, and loading/unloading. Workers’ radiation exposure during the process was systematically studied. A shielding package comprising a cylindrical and hexahedron structure was considered to reduce workers’ radiation exposure, and improved the treatment process’s efficiency. The shielding package’s effect on workers’ radiation exposure during the cutting and crushing process was also studied. The calculated annual radiation exposure of concrete treatment workers was below 1 mSv, which is the annual radiation exposure limit for members of the public. It was also found that workers involved in cutting and crushing were exposed the most.

Major accidents at nuclear power plants generate huge amounts of radioactive waste in a short period of time over a wide area outside the plant boundary. Therefore, extraordinary efforts are required for safe management of the waste. A well-established remediation plan including radioactive waste management that is prepared in advance will minimize the impact on the public and environment. In Korea, however, only limited plans exist to systematically manage this type of off-site radioactive waste generating event. In this study, we developed basic strategies for off-site radioactive waste management based on recommendations from the IAEA (International Atomic Energy Agency) and NCRP (National Council on Radiation Protection and Measurements), experiences from the Fukushima Daiichi accident in Japan, and a review of the national radioactive waste management system in Korea. These strategies included the assignment of roles and responsibilities, development of management methodologies, securement of storage capacities, preparation for the use of existing infrastructure, assurance of information transparency, and establishment of cooperative measures with international organizations.

RADTRAN is a code that assesses the radiation risk of radioactive material transportation. RADTRAN assumes that the package is a point source or a line source regardless of package type and corrects the external dose rate using a shape factor which depends on the critical dimension of the package. However, the external dose rate calculated using a shape factor may be different from the actual external dose rate. Therefore, it is necessary to analyze the effect of the shape factor on the external dose rate. In this study, the effect of the shape factor on the external dose rate in RADTRAN was analyzed by comparison with MCNP. This study analyzed change in external dose rate depending on the distance from the package and the critical dimension. The distance from the package was in the range of 1–800 m. The shape of the package was assumed to be cylindrical with a radius of 1 m, and the critical dimensions of the package were assumed to be 2, 4, and 8 m. Attenuation and build-up in the air were not considered to consider only the effect on the shape factor. When simulating the exposure situation using MCNP, the package was assumed to be a volume source, and flux by distance from the package was calculated using F5 tally. The dose rate at 1 m from the package was normalized to 2 mSv·hr−1. As a result of the analysis, the external dose rates of the package were higher in RADTRAN than in MCNP. For the critical dimension of 2, 4, and 8 m, when the distance from package is 1–10 m, the RADTRAN was 1.83, 4.08, and 5.27 times higher on average than MCNP, respectively. And when the distance from the package was 10–100 m and 100–800 m, RADTRAN was 1.10, 2.02, 3.01 times and 1.04, 1.92, 2.43 times higher than MCNP, respectively. It was found that the larger the distance from the package is and the smaller the critical dimension of the package is, the less conservatively RADTRAN assessed. It is because the shape of the package gets closer to the point source as the distance from the package increases, and the shape factor decreases as the critical dimension of the package decreases. The result of this study can be used as the basis for radiation risk assessment when transporting radioactive materials.

During the decommissioning of nuclear power plant (NPP), massive amount of concrete wastes is generated, which are non-radioactive and radioactive. The concrete is a representative construction material which affords reliable structural stability, good formability, and trustful integrity. Also, its reasonable neutron absorbing property allows the various application for many components, including building construction material, bio-shield concrete, etc. Due to the noted aspects of concrete, the radiological concrete characterization is classified as an important process for development of effective strategy for concrete management, in terms of process management and financial control during the decommissioning. The characterization of bio-shield concrete is important in waste management. The understanding and characterization of activation depth is essential for the determination of waste management strategy, classification of bio-shield concrete, and process development of decommissioning. On the other hand, concrete for construction application requires the evaluation of surface contamination of them. The concrete for containment building and its structure is rarely activated, but surface contaminated. In this paper, the reactor data from representative PWR reactors in the US is studied. The experience on Yankee Rowe, Maine Yankee, and Connecticut Yankee NPPs are systematically studied. The Yankee Rowe was a 4-loop PWR of Westinghouse design with 185 MWe. The Main Yankee was a 3- loop PWR of Combustion Engineering design with 864 MWe. The Connecticut Yankee was a 4-loop Westinghouse type with 560 MWe. The characterization studies on bio-shield concrete will be discussed in this paper, including activation depth, primary nuclides, etc.

The permanent shutdown of NPPs (nuclear power plants) has been growing steadily around the world. Also, permanent shutdown of old NPPs has been determined following to Kori-1 and Wolsong-1 in the Korea. Among issues of decommissioning of NPPs, especially, the management of radioactive waste is the most sensitive issue. According to IAEA, a large volume of radioactive concrete waste would be generated from decommissioning of nuclear facilities. Also, EC (European commission) expected that about 500 million tons of concrete will be produced in Europe by 2060 due to decommissioning of NPPs. It is known that the radioactive concrete consists of surface contaminated concrete and activated concrete. So, if contaminants from radioactive concretes can be removed using specific technologies, volume reduction of concrete can be achieved. Since there is no experience of decommissioning of NPPs in Korea, it is important to analyze previous cases. In this study, decontamination of radioactive concrete is analyzed through previous studies. Decontamination technologies of concrete are composed of mechanical methods, chemical methods, and thermal methods. Mechanical methods are physical technologies separating contaminants from concrete using scabbling, milling, and vacuum cleaning. In chemical methods, contaminants were removed from concrete using an oxidizing agent/reducing agent, acid/base. Thermal methods are removal technologies using lasers, microwaves, and pulsed power discharge. Some methods still have practical use cases, and further research is needed on the issue of generation of secondary waste. Review on the experience of decontamination of concrete show that waste of concrete generated during decommissioning of NPPs are expected to have effect of large volume reduction. However, many studies are needed because secondary waste and decontamination cost is sensitive issue of concrete generated during decommissioning of NPPs. In order to successful decommissioning of NPPs in Korea, various research of decontamination of concrete are need

HIC refers to a radwaste packaging container that can maintain integrity for more than 300 years in the general underground environment and disposal conditions in Korea. For HIC, the integrity of containers is verified according to the HIC regulation guideline for LLW and ILW disposal. Existing material tests include mechanical stability, permeability resistance, corrosion properties, chemical durability and biological resistance. In this study, a chemical durability test was conducted to prove the suitability of the HIC material by measuring the degree of chemical influence other than corrosion from the disposal environment. The chemical resistance evaluation method was used to simulate the disposal environment in the underground repository, and the amount of change in the physical properties of the degraded polymer concrete specimens according to the test time was confirmed. The technical standards considered leaching of material components, sulfation attack, acid attack, alkali, carbonate, and salt crystallization. The compressive strength and weight change of the specimens with time were checked by immersing them in a chemical solution that could leak major hazardous substances and wastes in the groundwater of the repository for several months. In addition, in order to evaluate the integrity in condition severe than the disposal environment, a flow was applied to a chemical solution having a concentration twice that of the basic chemical resistance test conditions, and the test period was extended twice to accelerate the deterioration of the specimen.

The Korea Nuclear Safety Act defines a high integrity container (HIC) as “a radioactive waste packaging container that can maintain its integrity for more than 300 years under the general underground environment and disposal conditions in Korea”, and detailed technical standards are not described. The US Nuclear Safety Commission’s “Low-Level Waste Licensing Branch Technical Position on Waste Form” describes the detailed requirements for solidification and HIC. The main contents of the US NRC technical position include limiting the free standing water, minimum design life, demonstrating mechanical, thermal and radiation stability, etc. In this study, the stability evaluation was performed to understand the mechanical strength with respect to horizontal and vertical loads. The basic property of polymer concrete was carefully evaluated, including compressive strength, structural fatigue resistance, etc. The long term creep test, loading of 40% of compressive strength, indicates that the polymer concrete exhibits good long term mechanical integrity.

After the Fukushima accident in 2011, a huge amount of radioactively contaminated water is being generated by cooling the melted fuel of units 1, 2 and 3. Most of contaminated water is seawater and underwater containing not only salt elements but also nuclear fission products with radioactivity. To treat the contaminated water, Cs/Sr removal facilities such as KURION and SARRY are being operated by TEPCO. Additionally, three ALPS facilities are on operation to meet the regularity standards for discharge to the sea. However, massive secondary wastes such as Zeolite, sludge and adsorbent is being generated by these facilities for liquid water treatment. The secondary wastes containing various radionuclide with Cs and Sr is difficult to store due to highly radioactive concentration and corrosive properties. In Japan, a variety of technologies such as GeoMelt vitrification, In-Can vitrification and CCIM vitrification is considered as a promising solution. In this study, they were reviewed, and the advantage and disadvantage of each technology were evaluated as the candidate technologies for thermal treatment of sludge radwaste.

In KHNP CRI, the 100 kW PTM (plasma torch melting) system was designed for the treatment and disposal technology of various radioactive wastes including the metal, concrete, liquid waste and insulator. The facility consists of melting chamber, thermal decomposition chamber, waste feeding system and off-gas treatment system. In this study, to evaluate the applicability of the PTM system, demonstration test was conducted using the radiation hazmat suit as combustible waste. The plasma melting chamber is pre-heated by 2nd combustion device and plasma torch for 5 hours. The temperature inside the plasma melting chamber is approximately 1,600°C. The combustible waste was put into the melting chamber by the pusher feeding device with the throughput of maximum 50 kg/hour. During the test, the power of plasma torch is 60–96 kW on the transferred mod. It was evaluated in terms of long-term integrity of PTM system on operation according to the waste throughput ratio.

Plasma melting technology has been considered as promising technology for treatment of radioactive wastes. According to the IAEA TECDOC-1527 report (2006), the technology has an advantage that it can treat regardless of waste types which is both combustible and non-combustible wastes. In particular, it is expected that a large amount of concrete, a representative non-combustible wastes, will be generated during the operation and dismantling of nuclear power plants. In order to treat the concrete waste in plasma torch melting system, various factors could be considered like the slag of electric conductivity, viscosity and melting temperature. Above all, as a critical factor, the viscosity of the melt is very important to easily discharge the melt. The viscosity of slag (SiO2-CaO-Al2O3 system) can be lowered by adding a basic oxide such as CaO, Na2O, MgO and MnO. The basic oxides are donors of oxygen ions. These oxides are called notwork breakers, because they destroy the network of SiO2 by reacting with it. In this study, the slag composition of the concrete waste was developed to apply the plasma torch melting. Also, demonstration test was performed with the developed slag composition and 100 kW plasma torch melting system.

During the operation or decommission of nuclear facilities, a large amount of dry active waste and cable waste with various shape and material is generated. Most of these wastes have almost no radioactive contamination and can be disposed of by incineration, landfill, recycling, etc. under clearance regulation. For clearance of radioactive waste, it is necessary to verify the characteristics of radiological contamination and prove that it meets the criteria for clearance regulation. According to the domestic clearance regulation, if it is difficult to measure radioactivity of wastes due to their surface condition using direct or indirect measurement methods, representative samples should be collected and analyzed for radioactivity. When sampling, it is desirable to collect samples of about 1 kg that can represent waste contamination per 200 kg or per 1 m2, and the homogeneity of the samples also should be demonstrated. However, in the case of dry active wastes, it is very difficult to prove the homogeneity of the samples because of surface shapes and conditions of the wastes. In particular, considering cable waste generated during the decommission, it is hardly capable to prove the representativeness of the sample, even though the inner shell of the covering material and the copper wire are almost uncontaminated. In this study, we show the development of a treatment system that makes it easy to prove the representativeness of samples when disposing of dry active waste or cable waste generated in nuclear facilities. The treatment device is designed in such a way that it has different storage unit and cutting unit suitable for the material characteristics of each waste type (soft, hard and cable), and therefore optimizes the efficiency of the shredding or cutting process. In addition, it is expected that the work efficiency in the radioactive treatment site with a narrow area can also be improved by providing a moving part on the device.

The safety assessment of a geological disposal system is performed over a period of hundreds of thousands of years, during which the activity of radionuclides in spent nuclear fuel decreases to natural radioactivity levels. During this period, the biosphere also experiences the long-term evolution of the surface environment including climate, terrain, and ecosystem changes. These changes cause changes in the water balance, which in turn change the pathways of radionuclides in the subsurface. Therefore, it is essential to consider these long-term changes in the surface environment for a reasonable biosphere safety assessment. For this purpose, this study developed the biosphere assessment module considering the long-term evolution of the surface environment, as a sub-module of APro (Adaptive process-based total system performance assessment framework). As a preceding study, the biosphere assessment module was previously developed using COMSOL for hydraulic and radionuclide transport processes, to simulate the pathway of radionuclides traveling from the shallow aquifer to the surface water body and soil. To consider the long-term evolution of the surface environment, the previous module needed to be improved to apply different water balances as boundary conditions of the module at each snapshot, which is a sub-time period divided based on the surface evolution data. To this end, this study utilized SWAT (Soil and Water Assessment Tool) which calculates the water balance using the surface environmental data including climate, terrain, land cover, and soil type. Conceptually, SWAT calculated annual water balance considering surface environmental changes, and certain components (i.e., groundwater recharge and hydraulic head of water bodies) of water balance were transferred to COMSOL as external data to simulate the pathway of radionuclide transport and spatio-temporal variability of radionuclides. At the current stage, the biosphere computational module has been developed to correspond to its conceptual model, and we plan to further test the applicability of the module using different surface environmental data.

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.

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.

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.