The mechanical safety of the container designed according to the IP-2 type technology standard was analyzed for the temporary storage and transportation of Very-Low-Level-Waste (VLLW) for liquid occurring at the nuclear facilities decommissioning site. The container was designed and manufactured as a composite shielding container with the effect of storing and shielding liquid radioactive waste using High Density Polyethylene (HDPE) and eco-friendly shielding material (BaSO4) with corrosion and chemical resistance. The main material of the composite shielding container is HDPE and BaSO4, the material of the cover, cage and pallet is SUS304, and the angle guard is elastic rubber. The test and analysis requirements were analyzed for structural analysis of container drop and lamination test. As test requirements for IP-2 type transport containers should be verified by performing drop and lamination tests. There should be no loss or dispersion of contents through the 1.2 m high free-fall drop and lamination test for a load five times the amount of transported material. ABAQUS/Explicit, a commercial finite element analysis program, was used for structural analysis of the drop and lamination test of the transport and storage container. (Drop test) It was confirmed that the container was most affected when it falls from a 45-degree slope. Although plastic deformation was observed at the edge axis of the cover, it was evaluated that the range of plastic deformation was limited to the cover and cage, and stress within the elastic limit occurred in the inner container. In the analysis results for other falling direction conditions, it was evaluated that stress within the elastic limit was generated in the inner container except for minor plastic deformation. In the case of on-site simulation evaluation, deformation of the inner container and frame due to the drop impact occurred, but leakage and loss of contents, which are major evaluation indicators, did not occur. (Lamination test) The maximum stress was calculated to be 19.9 MPa under the lamination condition for a load 5 times the container weight, and the maximum stress point appeared at the corner axis of the pallet. The calculated value for the maximum stress is about 10%, assuming the conservative yield strength of SUS304 is 200 MPa. It was evaluated that stress within the limit occurred. In the case of on-site simulation evaluation, it was confirmed that there was no container deformation or loss of contents due to the load.

Several previous simulation studies using various geochemical models have been carried out in several major analogue sites. The cases are beneficial when these studies provided the possibility of testing the geochemical models to be used to describe the migration of radionuclides in a future radioactive waste repository system. It was possible to interpret the complex transport behaviour of radionuclides such as uranium and thorium in an environment. We organize major natural analogue study sites from the previous literatures that provided information on the general geochemistry of the sites, in terms of groundwater composition and mineralogy. Also, we calculated aqueous speciation and the solid phases most likely to control their solubilities. The results obtained from the previous studies and this study vary depending on the tools used and on the conceptual models followed. Also, the results differed from the actual measured concentrations of trace metals or radionuclide analogues. The results obtained from these tests identify the main mathematical limitations of available geochemical models. However, the modelling results using a geochemical code with the thermodynamic database simulated well the observed behaviour of radionuclides, especially to identify the dominant processes controlling actinide mobilization and fixation. It was a useful outcome in terms of building confidence on the current geochemical tools to predict the concentrations of radionuclide analogues once the major geochemical characteristics were known. This study allows improving specific aspects of geochemical modelling using major natural analogue sites.

The radioactive waste disposal systems should consist of engineering and natural barriers that limit the leakage of radionuclide from spent nuclear fuel and fundamentally block groundwater from contact with radioactive waste. These considerations and criteria for designing a disposal system are important factors for the long-term stability evaluation of deep geological repository. Colloids and gases that may occur in the near-field and groundwater infiltrated from outside can be means to accelerate the behavior of radionuclide. The gas produced and infiltrated in the disposal system is highly mobile in the porous medium, and reactive gases in particular can affect the phase and behavior of radionuclide. A free gas phase (bubble) can be formed inside the canister if the partial pressure of the generated gas exceeds the hydrostatic pressure. If the gas pressure exceeds the critical endurance pressure of canister and buffer, then a gas bubble may push through the canister perforation and the buffer. It is also known that when gas bubbles are formed, radionuclide or colloids are adsorbed on the surface of the bubbles to enable accelerated movement. An experimental setup was designed to study the acceleration of nuclide behavior induced by gas-mediated transport. A high temperature and pressure reaction system that can simulate the deep disposal environment (500 m underground) was designed. It is also designed to install specimens to simulate gas flow in engineered barriers and natural barriers. The experimental scenario was set based on 1,000 years after the closure of the repository. According to the previous modeling results, the surface temperature of the canister is about 30 to 40 degrees and the gas pressure can be generated between the canister and the buffer is 5 MPa or more. In the experimental conditions, the saturation time of compacted bentonite was measured and the gas permeability of the compacted bentonite according to the dry density was also measured. Further studies are needed on the diffusion of dissolved gas into the compacted bentonite and the permeation phenomenon due to gas overpressure.

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.

Various diffusion experiments using geologic media have been carried out and it is often assumed that aqueous diffusion is the dominant transport mechanism. However, in some cases diffusive migration has been much faster than predicted in the model simulation. To explain such results surface diffusion of sorbing species was invoked. Experimental results were generally open to interpretation but possible existence of surface diffusion, whereby sorbed radionuclides could potentially migrate at much enhanced rates, necessitated investigation. The potential for surface diffusion of some sorbing nuclides on through-diffusion experiments using domestic rocks was examined. The apparent diffusion coefficients for sorbing cations were determined from their steady-state diffusion flux through rocks disks, while effective and pore diffusion coefficients were obtained with non-sorbing tracers through the same rocks. Diffusive transport models through domestic granites and granodiorites based only on pore diffusion did not often described adequately for sorbing cations. Thus, surface diffusion should be considered. Then what was the most important measure to estimate surface diffusion? As far as we examine, the sorption reversibility provides a hint of surface diffusion. The reversible sorbing species, for example, Sr, has a remarkable surface diffusion contribution, whereas surface diffusion has a relatively small contribution for irreversibly sorbing species such as Cs and Am under domestic experimental conditions.

Some Spent Fuel Pools (SFPs) will be full of Spent Nuclear Fuels (SNFs) within several years. Because of this reason, transporting the SNF from SFP to interim storage facilities or permanent disposal facilities should be considered. There are two ways to transport the SNF from a site to other site, one is the land transportation with truck or train, and the other is the maritime transportation with ship. The maritime transportation has some advantages compared with the land transportation. The maritime transportation method uses safer route which is far from populated area than land transportation method, and transport more weight than land transportation method. However, the cask should be loaded into the ship for the maritime transportation, and there is a possibility of a drop accident of the cask onto the ship. Therefore, it is necessary to evaluate the structural integrity of the cask and ship for the drop accident during the loading process. To evaluate the structural integrity of the cask and ship, it is necessary to determine the analysis conditions that caused the greatest damage in the drop accident. There may be various conditions such as the drop angle of the cask, the initial falling speed, the drop position onto the ship, the size of the ship, etc. This study set the drop angle of the cask and the drop position onto the ship as the simulation variables, which have high possibility to occur during cask drop. However, the others are excluded since they are controllable by worker. In this paper, various drop angle (0, 15, 30, 45, and 70 degree) of the cask were simulated to define the greatest damage condition. KORAD-21 cask model was used for Finite Element Analysis (FEA), and FEA was performed to simulate a horizontal drop (1 m drop). The strain-hardening material properties for the deck were used as HT36 steel. The Cowper-Symonds constitutive model for HT36 was used to consider the strain rate effect. A Tie-down structure for supporting the cask was modeled with the cask model which contained inner structures like canister, basket, etc. Structural integrity of the cask and tie-down structure were evaluated using the von-Mises stress and equivalent plastic strain (PEEQ), and one of the ship deck was evaluated using deflection of ship deck and equivalent plastic strain. Compared with each cask drop angle conditions, 45 degree of the cask drop angle showed the highest deflection and PEEQ values, but did not exceed ultimate strain of HT36. In the ship deck, the corner of deck showed the highest PEEQ value in all simulation cases. As the result, the 45 degree of the cask drop angle condition results was more conservative than other conditions, and the corners of deck failure was able to evaluate ship safety.

This paper intends to present considerations on the question of what is the “load standard” or “design load” for integrity evaluation under normal transportation conditions and what type of design load is good for users. This suggests a direction for subsequent research on producing design loads that transport business companies can utilize without difficulty. Several studies have been conducted to evaluate the integrity of spent nuclear fuel during normal transportation. A representative study recently conducted is the Multi-modal Transportation Test (MMTT) conducted using a commercial spent nuclear fuel cask by US DOE in 2017. In Korea, additional transport tests were planned to acquire sufficient test data under the conditions of road and sea transport considering the Korean situation. As a result, road transport tests were carried out in 2020 and sea transport tests were carried out in 2021. In the road transport test, a driving test that simulates various road conditions and a test that cycled a 4.5 km road eight times were performed. In most cases, the maximum acceleration of less than 1 g occurred, and the maximum strain was less than 48 με. For the sea transport test, the magnitude of both the maximum acceleration and the maximum strain were lower than those in the road transport test. We concluded tentatively that the integrity of spent fuel under normal conditions of transport was satisfactory with a large margin. However, when the storage business is realized and the transport of spent fuel becomes visible, the storage and transport business companies will have to prove the maintenance of the integrity of the spent fuel under normal transport conditions at the request of the regulatory agency. The transport business companies can transport the spent nuclear fuel by using different types of transport casks and different types of trucks and ships from those used in the tests mentioned above. However, it is absurd to have to prove the integrity of spent nuclear fuel by performing expensive tests again. Therefore, in this study, the design load that can be used by transport business companies is to be presented. The design load to be presented should satisfy the following requirements. The design load should be applicable including some differences in the transport cask or transport system, or different design loads should be presented according to the differences. The location where this design load is applied is to be specified (e.g. fuel rod, basket, internal structure). Requirements according to the operating speed of the transport system should be presented together. The type of design load is to be presented (e.g. PSD, SRS, FDS etc.). Other types of standards may be presented. For example, a speed limit for a vehicle carrying spent nuclear fuel may be suggested, or a speed limit for a vehicle passing through a speed bump may be suggested. In order to present such a reliable design load, a multi-axis vibration excitation shaker table test will be carried out. Though this shaker table test, the behavior of the nuclear fuel assembly is closely evaluated by applying the data obtained from the road and sea transport tests previously performed as an input load. In addition, FDS (Fatigue Damage Spectrum) will be produced and applied to experimentally evaluate the durability of fuel assemblies under normal transport conditions.

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.

Currently, the HI-STAR 63 transport cask, developed to transport CANDU spent nuclear fuel from the wet storage pool to the dry storage facility which is called the MACSTOR/KN-400, has a transport capacity of 120 bundles, which is unfavorable when considering transportation costs and other related aspects. According to the ‘Basic Plan for High-Level Radioactive Waste Management (draft)’, the total amount of CANDU spent nuclear fuel is expected to be approximately 660,000 bundles. To safely and efficiently transport this amount to interim storage facilities, it is essential to develop a large-capacity transport cask. Therefore, we have been developing a large-capacity PHWR spent nuclear fuel transport cask, called the KTC-360 transport cask. According to the transport-cask related regulations, the KTC-360 transport cask was classified as a Type B package, and such packages need to maintain integrity under the normal transport and accident conditions described in these regulations. To prove the thermal integrity of this cask under the normal transport and accident conditions, high-temperature and fire tests were performed using a one-third slice model of an actual KTC-360 cask. The results revealed that the surface temperature of the cask was 62°C, indicating that such casks need to be transported exclusively. The highest temperature of the CANDU spent nuclear fuel was predicted to be lower than the melting temperature of Zircaloy-4, which was the sheath material used. Therefore, if normal operating conditions are applied, the thermal integrity of a KTC- 360 cask could be maintained under normal transport conditions. The fire test revealed that the maximum temperatures of the structural materials, stainless steel, and carbon steel, were 446°C lower than the permitted maximum temperatures, proving the thermal integrity of the cask under fireaccident conditions.

2020년 1월 23일 이후 중국에서 신종 코로나바이러스 감염증(COVID-19)으로 인한 봉쇄 조치가 전국으로 확대 되고 있었다. 그러나, 한국에서는 2020년 2월 1-2일에 PM10 질량농도 일평균 최대 88-98 μg m−3의 고농도 연무가 발생하였다. 이 기간에 동아시아 지역은 850 hPa 기온 아노말리가 양(+), 동서류 아노말리는 음(-)으로 온난하고 정체적인 기단의 영향을 받고 있었다. 동아시아 지역의 인위적 배출량 감소에 따른 한국의 PM10 장거리 수송의 영향을 분석하기 위하여 WRF-Chem을 활용하였다. WRF-Chem에 인위적 배출량을 변화 없이 적용한 BASE와 인위적 배출량을 50%로 감소시켜 적용한 CTL의 PM10을 한국의 지상 측정값과 민감도 분석을 수행하였다. CTL에서 PM10의 IOA는 0.71로 BASE의 0.67보다 높게 나타났다. 이것은 중국의 COVID-19 봉쇄 조치로 인해 인위적 배출량이 감소한 것으로 분석된 다. 또한, 한국 이외의 지역 배출량을 0으로 설정한 BASE_ZEOK와 CTL_ZEOK를 모의하여 BASE와 CTL에서의 장 거리 수송 기여도의 변동을 분석하였다. CTL은 BASE와 비교하여 배출량이 50%로 감소하였지만 PM10 장거리 수송 기여도는 10-20% 감소한 것으로 나타났다. 동아시아 지역의 배출량 감소에 따라 풍하측 한국의 PM10 장거리 수송 기 여도 변동이 선형적으로 반응하지 않는 것은 종관 기상 변동이 영향을 주는 것으로 보인다. 2월 1-2일 한국의 고농도 PM10 연무 사례에 대한 CTL에서 PM10 에어로졸 성분의 장거리 수송 기여도는 기타 무기물이 80-90%로 가장 높았고, 질산염은 30-60%, 황산염은 0-20%, 암모늄은 30-60%를 나타내고 있었다. 중국의 봉쇄 조치로 인하여 교통 및 물류 수 송이 감소하면서 2차 에어로졸이 감소한 것으로 보인다.

The structural safety of prototype transport and storage containers for very low-level radioactive liquid waste was experimentally estimated for its localization development. Transport containers for radioactive liquid waste have been researched and developed, however, there are no standardized commercial containers for very low-level radioactive waste in Korea. In this study, the structural safety of the designated IP-2 type container capable of transporting and temporarily storing large amounts of very low-level liquid waste, which is generated during the operation and decommissioning of nuclear power plants, was demonstrated. The stacking and drop tests, which were conducted to determine the structural integrity of the container, verified that there was no external leakage of the contents in spite of its structural deformation due to the drop impact. This study shows the effort required for the localization of the technology used in manufacturing transport and storage containers for very low-level radioactive liquid waste, and the additional structural reinforcement of the container in which the commercial intermediate bulk container (IBC) external frame was coupled.

최근 동아시아 지역에서 인위적 배출량의 감소에도 불구하고, 봄철에 한국에서는 잦은 연무 사례가 발생하고 있 다. 북동 태평양에서 자주 발생하는 대기 블로킹은 지구 규모 대기 변동과 동아시아 지역의 서풍 기류를 정체시키기도 한다. 2019년 3월 동아시아 지역의 온난하고 정체적인 종관 기상 특성이 알래스카 대기 블로킹이 발생한 6-7일 후에 일어나고 있었다. 특히, 2019년 3월 18-24일에 발생한 알래스카 대기 블로킹은 3월 25-28일 동안 한국에서 일평균 미 세먼지(particulate matter; PM10) 질량농도가 50 μg m−3을 넘는 고농도 PM10 연무 사례가 발생하는 데 영향을 미치고 있 었다. 한편, WRF-Chem 모델을 활용하여 한국의 고농도 PM10 연무 사례에 대한 인위적 배출의 장거리 수송 기여도는 30-40%를 나타내고 있었다. PM10 에어로졸 구성 성분인 황산염, 질산염, 암모늄, 블랙 카본, 유기 탄소, 기타 무기물의 장거리 수송 기여도는 각각 10-15, 20-25, 5-10, 5-10, 5-10, 15-20%를 나타내었다. 질소 산화물이 온난하고 정체적인 대기에서 암모늄과의 광화학 반응으로 형성된 질산암모늄은 한국의 고농도 PM10 연무 사례에 대한 장거리 수송 기여도 가 PM10 에어로졸 중 가장 큰 비중을 나타내고 있었다.

PURPOSES : The logistic roads for freight transport along to the new port of Busan have been suffered by the rapid weather changes including high temperature and torrential rain. As a result, the roads require annual repair, which have been distressed seriously by the heavy logistic and environmental loads. Therefore, we need to identify the cause of the road pavement distresses and find a proper design method to minimize the pavement distress in order to prohibit the problem aggravated. METHODS : The damaged conditions of the logistic roads were investigated on-site. In addition, applied pavement designs, real traffic volumes, and historical climatic information were intensively collected for this project. With the investigated and collected data Korean pavement design program (KPRP) was implemented to analyzed the causes of the damaged roads and conceive the pavement design draft optimized for the roads. RESULTS : According to the investigation and KPRP analysis, the traffic volume to transport freights impacts significantly the pavement distress, so that a higher PG grade binder type should be used, for which polymer modified asphalt (PMA) binders are recommended. Moreover, its pavement thickness should be increased to secure load bearing capacity, but thickening the pavement has been discouraged due to difficulties induced by the road-sectional change, especially road-height change. CONCLUSIONS : In conclusion, 5cm PMA overlay is suggested for the normal-scale maintenance, and 7cm PMA overlay for large-scale maintenance. Besides these, the application of Polymer-modified Stone Matrix Asphalt (PSMA) using PG76-22 binder would be the best preventive maintenance method, which has been well know as having higher fatigue resistant performance than general PMA. However, if we use PSMA, quality control should be very cautious since PSMA can be very susceptible premature distress if its production and construction are improperly proceeded.