The U.S. Nuclear Regulatory Commission (NRC) states that every environmental report prepared for the licensing stage of a Pressurized Water Reactor shall contain a statement concerning risk during the transportation of fuel and radioactive wastes to and from the reactor. Thus, the licensee should ensure that the radiological effect in accidents, as well as normal conditions in transport, do not exceed certain criteria or be small if cannot be numerically quantified. These are specified in 10 CFR Part 51 and applied in NUREG-1555 Supplement 1 Revision 1, which deals with Environmental Standard Review Plan. Corresponding regulations in Korea would be the Nuclear Safety and Security Commission Notice No. 2020-7. In Appendix 2 of the Notice, guides on the radiological environmental report for production and utilization facilities, spent nuclear fuel interim storage facilities, and radioactive waste disposal facilities. In this guide, unlike the regulations in the U.S., there are no obligations for radiological dose assessment for workers and public during the transportation. Therefore, overall regulations and their legal basis related to risk assessment during transportation conducted for the environmental report in the U.S. were analyzed in this study. On top of that, through the comparison with regulations in Korea, differences between the two systems were figured out. Finally, this study aims to find the points in terms of assessing transport risk to be revised in the current regulatory system in Korea.

Low- and intermediate-level radioactive wastes have been disposed of in the first-phase deep underground silo disposal at Gyeongju in South Korea. These radioactive wastes contain harmful radionuclides such as Uranium-238 (238U), which can pose long-term and deleterious effects on humans and the natural environment. Ethylenediaminetetraacetic acid and isosaccharinic acid, which can be formed via cellulosic waste degradation under high alkaline conditions might considerably enhance the transport behavior of 238U with the intrusion of rainwater and groundwater. In this study, the engineered barriers (concrete and grout) and natural barriers (sedimentary rock and granite) were used to investigate the 238U transport behavior in artificial cementitious porewater of State I (pH 13.3) and State II (pH 12.5) based on groundwater or rainwater. The surface properties and geochemical compositions of barrier samples were characterized using XRD, XRF, SEM-EDX, and BET. The transport behaviors of 238U in various solution conditions were observed by sorption distribution coefficient (Kd) at a range of initial chelating agents concentration (10-5-10-2 M). The sorption behavior of 238U was retarded more in the engineered rock barriers than in the natural rock barriers. The mobility enhancement of 238U was more significant in State I than in State II. In comparison with the absence of chelating agents, negligible changes in the Kd values of 238U were observed at less than initial chelating agent concentrations of 10-4 M. However, the Kd values of 238U were significantly reduced at initial chelating agent concentrations higher than 10-3 M. Therefore, these experimental findings show that the transport behavior of 238U into the geo- and bio-sphere could be accelerated by the presence of chelating agents and the type of cement degradation states.

The “shadow zone” is defined as a region below a flow obstacle, such as a vault, in unsaturated soils. Due to the capillary discontinuity of the cavity, water saturation on the top and side of the cavity is higher than the ambient saturation. On the bottom of the cavity, however, there is a region where water saturation is lower than ambient saturation. Undoubtedly, a shadow zone may also exist below a LILW disposal vault built in subsurface soils above the water table before the vault is fully degraded. During the degradation, flow in the shadow zone is controlled by the rate of water infiltrating the degrading vault. In this study, as one of the efforts to be made for enhancing safety margin by a realistic safety assessment of the engineered vault type LILW disposal facility, the shadow zone effect is investigated by a numerical parametric study using AMBER code. The conceptual model and data were excerpted from IAEA, ISAM Vault Test Case for the liquid release design scenario. It is assumed that the nearfield barriers degrade with time. In order to compare a visible shadow zone effect, the vault degradation period is assumed to be both 500 and 1,000 years, and the shadow zone depth to be varied according to unsaturated zone lithology. It can be seen that with a shorter shadow zone (2.7 m), radionuclides arrive at the water table earlier than with a full shadow zone (55 m) due to increased advection rate in the unsaturated zone. This effect tends to be more visible in the case of a longer degradation period. For radionuclides with short residence time relative to their half-lives in the unsaturated zone, such as Tc-99 and I-129, the radionuclides are shown to come out because they will arrive sooner, thereby allowing less peak release rate, when the shadow zone effect is considered. Once the vault is completely degraded and the infiltration rate of water flowing through the vault is equal to the ambient rate, the shadow zone effect disappears. In this example calculations using IAEA ISAM Vault Test Case input parameters, it might not be shown a significant shadow zone effect. Nevertheless, when the extent of the shadow zone is determined through more sophisticated hydraulic studies in the unsaturated soils surrounding the vault, the shadow zone effect would be checked up on the realistic near-field radionuclide transport modeling in order to contribute to gaining safety margins for post-closure safety assessment of the Wolsong 2nd phase LILW disposal facility.

Excavation Damaged Zone (EDZ) is created by the excavation of deposition holes and disposal tunnels at high-level radioactive waste repository that causes macro- and micro-fracturing in the surrounding rock. Since EDZ can significantly increase the hydraulic transmissivity in the rock and act as a major pathway of leaked radionuclides, consideration of EDZ in terms of safety assessment is very important. Moreover, long-term stress changes such as stress redistribution due to excavation of nearby deposition holes and disposal tunnels, thermal stress due to temperature rise, effective stress change due to pore pressure change, and swelling pressure of bentonite buffer can increase EDZ size and change in thermal-hydraulic-mechanical properties, and consequently, it can affect the transport of radionuclides. Therefore, in order to analyze the effect of long-term evolution of EDZ on radionuclide transport, it is essential to conduct numerical analysis considering the coupled Thermal-Hydraulic- Mechanical (THM) behavior in EDZ. In order to simulate the behavior of EDZ, coupled THM model was developed using the Adaptive Process-based total system performance assessment framework for a geological disposal system (APro) proposed by the Korea Atomic Energy Research Institute (KAERI). The concept of damage was introduced to demonstrate the jointed rock as a continuous medium. Among several damage models, Mazars damage model was applied in this study. Mazars damage model is the most well-known model for concrete which has similar behavior with rock as brittle material, and the input data of the model can be easily obtained through laboratory testing. If damage occurs due to the influence of thermal-hydraulic-mechanical coupled behavior at the bedrock, the properties change according to the degree of damage, and as a result, the migration of the radionuclide is affected. Based on this conceptual model, radionuclide transport model in the near field considering the long-term evolution of EDZ was developed. To investigate the effect of EDZ in terms of process-based performance assessment, the modeling results with and without EDZ were compared. Finally, by simulating the coupled THM behavior of EDZ with damage model, the effect of long-term evolution of EDZ on radionuclide transport was investigated.

The saturation rates of the spent fuel (SF) wet storage at the Kori Nuclear Power Plant (NPP), Hanbit, and Hanul are 83.3%, 74.2%, and 80.8% as of the fourth quarter of 2021. The storages of Kori NPP and Hanbit NPP are expected to be saturated in 2031, and Hanul is expected to be saturated in 2032. Therefore, the construction of an interim storage facility to store the SF temporarily stored in the NPP was planned, and preparations for the safe transport of the SF are required. In this paper, radiological preliminary assessment using NRC-RADTRAN in the process of sea transport of SF from the wet storage or ISFSI of the Hanbit NPP to the optional interim storage facility was performed. Since domestic SF transport vessels are not currently in operation, the specifications of the UK Pacific Grebe vessel which can carry up to 20 casks were used. The transport cask used the specifications of KORAD-21, a transport container developed in Korea. Because it can carry more SF assemblies than the existing KN-18. In addition, a land transport safety test was conducted in 2020 and a sea transport test is planned. The sea transport route was entered by referring to the transport route of domestic low and intermediate level waste. The accidents rate was calculated using statistics on maritime accidents from 2017 to 2021. The probability accidents along the transportation route were evaluated as 3.152E -10. When transporting to an interim storage facility, the SF expected to be the main transport target was selected as WH 17X17, combustion 45,000 MWD/MTU, and concentration of 4.5%. The source term was calculated and entered according to this data and the release fraction was entered with reference to the DOE report. In the case of normal transport without accident, the individual dose of the crew member and public residents were estimated to be 0.0525% and 0.000492% of the annual limit of 1 mSv/yr for the general public. Under the accident conditions, the annual individual doses of residents were 0.0011%, 0.0023%, 0.0034%, and 0.0046% of the annual limit of 1 mSv/yr when carrying 5, 10, 15, and 20 casks. Currently, the site of the interim storage facility has not been precisely determined, but a preliminary radiation assessment through sea transport resulted in a significantly lower than the limit. Combined scenario sea transport followed by land transport will be carried out in the next stage of study.

In case a spent nuclear fuel transport cask is lost in the sea due to an accident during maritime transport, it is necessary to evaluate the critical depth by which the pressure resistance of the cask is maintained. A licensed type B package should maintain the integrity of containment boundary under water up to 200 m of depth. However, if the cask is damaged during accidents of severity excessing those of design basis accidents, or it is submerged in a sea deeper than 200 m, detailed analyses should be performed to evaluated the condition of the cask and possible scenarios for the release of radioactive contents contained in the cask. In this work, models to evaluate pressure resistance of an undamaged cask in the deep sea are developed and coded into a computer module. To ensure the reliability of the models and to maintain enough flexibility to account for a variety of input conditions, models in three different fidelities are utilized. A very sophisticated finite element analysis model is constructed to provide accurate response of containment boundary against external pressure. A simplified finite element model which can be easily generated with parameters derived from the dimensions and material properties of the cask. Lastly, mathematical formulas based on the shell theory are utilized to evaluate the stress and strain of cask body, lid and the bolts. The models in mathematical formula will be coded into computer model once they show good agreement with the other two model with much higher fidelity. The evaluation of the cask was largely divided into the lid, body, and bottom, bolts of the cask. It was confirmed that the internal stress of the cask was increased in accordance with the hydrostatic pressure. In particular, the lid and bottom have a circular plate shape and showed a similar deformation pattern with deflection at the center. The maximum stress occurred where the lid was in the center and the bottom was in contact with the body. Because the body was simplified and evaluated as a cylinder, only simple compression without torsion and bending was observed. The maximum stress occurred in the tangential direction from the inner side of the cylinder. The bolt connecting the lid and the body was subjected to both bending and tension at the same time, and the maximum stress was evaluated considering both tension and bending loads. In general, the results calculated by the formulas were evaluated to have higher maximum stresses than the analysis results of the simplified model. The results of the maximum stress evaluation in this study confirms that the mathematical models provide conservative results than the finite element models and can be used in the computer module.

This study is to investigate fuel cladding temperature in a transport system for the purpose of developing a methodology for evaluating the thermal performance of spent fuel. Detailed temperature analysis in the transport system is important because the degradation mechanism of the fuel cladding is generally sensitive to temperature and temperature history. In such a system, the magnitude of the temperature change is determined by examining the temperature sensitivity of fuel assemblies and system components including fuel cladding temperature, considering the material properties, component specifications, component aging mechanism, and heat transfer mechanism. The sensitivity analysis is performed using heat transfer models by computational fluid dynamics for the horizontal transport system. The heat transfer within the system by convection, conduction and thermal radiation is calculated by thermal-hydraulic analysis code FLUENT. The calculation region is divided into a basket cell and a transport cask. The thermal analysis of the basket cell is for predicting the fuel cladding temperature. And the reason for analyzing the transport cask is to provide the boundary condition for the basket cell by reflecting the external environmental conditions. Here, the basket cell containing the spent fuel assembly is modeled on the homogeneous effective thermal conductivity. The purpose of this analysis is to evaluate fuel cladding temperatures for the following four main items. That is the effect of surface emissivity changes in basket due to the oxide layer of the fuel cladding, the effect of degradation of the canister backfill helium gas, the effect of fuel assembly position in basket cell on fuel cladding and basket temperatures in canister, and the effect of using the homogeneous effective thermal conductivity model instead of the fuel assembly in basket cell. As a result of the analysis, the maximum temperatures in basket cells are evaluated for the above four items. Thermal margins for each item are investigated for thermal performance requirements (e.g., peak clad temperature below 400oC).

Thermal analysis and safety assessment of spent fuel transport cask are mainly conducted using commercial Computational Fluid Dynamics (CFD) codes based on Finite Volume Method (FVM). The reliability and predictability of CFD codes have greatly been improved by the development in the computer systems, and are widely used to calculate heat flow in complex structures that cannot be analyzed theoretically. In the field of thermal analysis using the CFD code, it is important to clearly reflect the physical model of the transport cask, and a grid configuration suitable for the physical model is essential for accurate analysis. However, since there are no clear standard and guidelines for grid configuration and size, it is highly dependent on the user’s insight. Spatial discretization errors result from the use of finite-width grids and the approximation of the differential terms in the model equations by difference operators. Since the user usually cannot change the truncation error order of a given discretization scheme, spatial discretization errors can only be influenced by the provision of optimal grids. Therefore, it is necessary to quantify the spatial discretization errors caused by the grid. In the case of Orano TN’s NUHOMS® MP197 transport cask, considering four grids for two sets, the temperature uncertainty of the neutron shield, which has the lowest margin at the limit temperature among transport cask components, was quantified by applying 5-step procedure of the Grid Convergence Index (GCI) method for the uncertainty estimation presented in ASME V&V 20-2009. In the case of domestic spent nuclear fuel transport cask (KORAD21), neutron shield among the transport cask components has the lowest margin at the limited temperature. Accordingly, in this study, the temperature uncertainty of the neutron shield was quantified by applying GCI to three sets considering seven grids. As a result of the calculation, the uncertainty was less than ± 1°C, and the temperature of the neutron shield including the uncertainty was evaluated to be maintained below the limit temperature of 148°C.

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 must be able to withstand a temperature of 800°C for a period of 30 min. It is desirable to conduct a test using a fullscale model of a shipping package when performing tests to evaluate its integrity. However, it is costly to perform a test using a full-scale model. Therefore, to evaluate the thermal integrity of the KTC-360 transport cask, the fire test was conducted using a slice model. For comparison purposes, the fire test was also carried out using a 1/4 scale model. In the fire test using a slice model and in the fire test using a 1/4 scale model, the maximum temperature of the cask body was lower than the permitted maximum temperature limit. Therefore, the thermal integrity of the KTC-360 transport cask could be considered to be maintained. The temperature results from the fire test using a slice model were higher than those of the fire test using a 1/4 scale model. Therefore, the effect of flame on a transport cask without combustible materials, such as the KTC-360 transport cask, seems to be affected by the reduction in the time rather than the size reduction.

As the saturation rate of temporary storage facilities for spent nuclear fuel increases, regulatory demands such as interim storage and permanent disposal of spent nuclear fuel are expected to begin in earnest. Considering the domestic situation where all nuclear power plants are located on the waterfront site, the interim storage site is also likely to be located on the waterfront site, and maritime transportation is one of the essential management stages. Currently, there are no independently developed maritime transportation risk assessment code in Korea, and no research has been conducted to evaluate the release of radioactive waste due to the sinking of transport container. Therefore, it is necessary to secure technology to properly reflect the domestic maritime transportation environment and to assess the impact of the sinking accident and to carry out safety regulations. To accurately calculate the releaser rate of radionuclides contained in a cask with breached containment boundary, the flow rate through the gap generated in the containment boundary should be calculated. The fluid flow through this gap which is probably in micro scale in most situations should be evaluated combining the fluid flow inside and outside the cask. In this study, a detailed computational fluid dynamics model to evaluate the internal fluid flow in the cask and a simplified model to capture the fluid flow and the heat transfer around the cask in the sea are constructed. The results for the large scale model are compared with the analytic formula for verification of heat transfer coefficient and they showed good agreements. The heat transfer coefficient thus found can be used in the detailed model to provide more realistic data than those obtained from assumed heat transfer coefficient around the surface of the cask. In the future, fluid flow through the gap between the lid and the body of the cask will be evaluated coupling the models developed in this work.

The safety of a KTC-360 transport cask, a large-capacity pressurized heavy-water reactor transport cask that transports CANDU spent nuclear fuel discharged from the reactor after burning in a pressurized heavy-water reactor, must be demonstrated under the normal transport and accident conditions specified under transport cask regulations. To confirm the thermal integrity of this cask under 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 must be transported separately. 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 can 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 fire accident conditions.

The transport of radionuclides at oceanic scales can be assessed using a Lagrangian model. In this review an application of such a model to the Atlantic, Indian and Pacific oceans is described. The transport model, which is fed with water currents provided by global ocean circulation models, includes advection by three–dimensional currents, turbulent mixing, radioactive decay and adsorption/release of radionuclides between water and bed sediments. Adsorption/release processes are described by means of a dynamic model based upon kinetic transfer coefficients. A stochastic method is used to solve turbulent mixing, decay and water/sediment interactions. The main results of these oceanic radionuclide transport studies are summarized in this paper. Particularly, the potential leakage of 137Cs from dumped nuclear wastes in the north Atlantic region was studied. Furthermore, hypothetical accidents, similar in magnitude to the Fukushima accident, were simulated for nuclear power plants located around the Indian Ocean coastlines. Finally, the transport of radionuclides resulting from the release of stored water, which was used to cool reactors after the Fukushima accident, was analyzed in the Pacific Ocean.

This study compares the characteristics of a compact TiO2 (c-TiO2) powdery film, which is used as the electron transport layer (ETL) of perovskite solar cells, based on the manufacturing method. Additionally, its efficiency is measured by applying it to a carbon electrode solar cell. Spin-coating and spray methods are compared, and spraybased c-TiO2 exhibits superior optical properties. Furthermore, surface analysis by scanning electron microscopy (SEM) and atomic force microscopy (AFM) exhibits the excellent surface properties of spray-based TiO2. The photoelectric conversion efficiency (PCE) is 14.31% when applied to planar perovskite solar cells based on metal electrodes. Finally, carbon nanotube (CNT) film electrode-based solar cells exhibits a 76% PCE compared with that of metal electrodebased solar cells, providing the possibility of commercialization.

The Co-60 is a radioactive material widely used in domestic and foreign medical, industrial, health and research fields. Currently, world market for the Co-60 is about 80 MCi/yr and is expected to grow to 150 MCi/yr by 2025. For the Co-60, Nordion of Canada occupies about 80% of the world market. In the case of Korea, a small amount of sources with relatively low radioactivity intensity are produced using research reactors, but most of the Co-60 is entirely dependent on imports. Accordingly, although the technical feasibility of the Co-60 production technology using the PHWR was evaluated, it was evaluated as a negative result on the additional construction of a hot cell, core management, safety analysis and economic feasibility. Canada, the main producer of the Co-60, is also conducting research on the Co-60 production technology using PWR with GE-Hitachi and Westinghouse as the number of PHWR is expected to decrease. In Korea, it is necessary to preoccupy the Co-60 production technology and auxiliary technology using the PWR by utilizing excellent technology, and active research is being conducted to secure unique nuclear power technology that does not depend on foreign countries. Therefore, in this study, the thickness and weight of the radioactive shielding required for handling (transport) of Co-60 produced using the PWR were calculated.

To rationalize the protection of spent nuclear fuel transport storage cask, we intend to investigate the status of domestic and foreign safety regulations and related technologies to develop sabotage scenarios and analyze the protection performance and radiation impact of transport storage cask. It is essential to conduct an aircraft collision safety evaluation on spent nuclear fuel transportation and storage casks in Korea due to changes in laws and regulations related to nuclear power plant design and demand for enhanced safety. Domestic and foreign research on the protection performance of spent nuclear fuel transport storage cask was based on 9.11 events, and the results of all studies show that the speed of the aircraft and leakage of nuclear materials are insignificant. The Sandia National Laboratory (SNL) calculates Aerosol emissions from spent fuel damage in the event of sabotage and calculates Source Term based on the Durbin-Luna model. In this paper, radiation sensitivity analysis was performed due to damage to the carrier according to the size of the accident, assuming that there was a hole enough to basket from the external shell among the collision scenarios identified for domestic cask models.

Radioactive waste generated during the decommissioning of Kori Unit 1 can be packaged in a transport container under development and transported to a disposal facility by sea transport or land transport. In this study, the cost of each transport method was evaluated by considering the methods of land transport, sea transport, and parallel transport of the radioactive waste dismantled at Kori Unit 1. In evaluating the shipping cost, the shipping cost was evaluated by assuming the construction of a new ship without considering shipping by CHEONG JEONG NURI, which is currently carrying operational waste. Since the cargo hold of CHEONG JEONG NURI was built to fit the existing operating waste transport container and is not suitable for transporting the transport container currently under development, sea transport using CHEONG JEONG NURI was excluded in this paper. In the case of on-road transportation, the final fare for each distance was calculated in accordance with the Enforcement Decree of the Freight Vehicle Transportation Business Act, and the cost of onroad transportation was evaluated by estimating the labor cost of the input manpower required for onroad transportation. The cost of on-road transportation was estimated to be approximately KRW 510 million, the product of the total number of transports 459 times the sum of the cost of transportation vehicle freight cost of about KRW 720,000 and the labor cost of input personnel of KRW 380,000. It is difficult to predict the cost of building a new ship at this point, as the cost of building new ship is determined by the cost of number of items such as ship design specifications and material prices, labor costs, and finance costs at the time of construction. Accordingly, considering the 2% annual inflation rate based on the shipbuilding cost (about KRW 26 billion) and financing cost (about KRW 12 billion) at the time of construction of the CHEONG JEONG NURI (2005 yr.), decommissioning of Kori Unit 1 (2025 yr.) construction cost finance cost was estimated and evaluated. According to the result of comparing the transport cost for each transport scenario, land transport is about 510 million won, which is advantageous in terms of economic feasibility compared to the sea transport scenario. However, when transporting by land, it is disadvantageous in terms of acceptability of residents because it is transported multiple times on general roads. The cost of building a new ship is about KRW 56.4 billion, which is disadvantageous in terms of the cost of transporting waste from the dismantling of Kori Unit 1. But, in the future, cost reduction can be expected if waste materials issued when dismantling nuclear power plants are transported.

For transport containers for radioactive wastes, a drop test should be performed at a height of 0.3– 1.2 m on a rigid target depending on the weight as a normal condition in the regulation. In the drop test, a strain gauge is commonly used to measure the local strain, and the position of the strain gauges is determined by the experiences of the engineer in advance of the test. For this reason, the strains can be measured at only predetermined points. The DIC (Digital Image Correlation) method using highspeed cameras can be used to measure the change in strain over the region of interest. In addition, it is possible to measure effectively even in areas with high strain gradients that are difficult to measure with strain gauges. Therefore, the DIC method can measure the strain change according to time over the entire load path. When the drop test of the transport container is performed, the impact load is delivered through the lower corner fittings-corner posts-upper corner fittings-lids. In this study, white spray was sprayed on these main load path, and black speckles were created on the spayed surface to trace the rigid motion of speckles. The images taken during the drop test can be used to create a strain field over region of interest.