Because most spent nuclear fuel storage casks have been designed for low burnup fuel, a safety-significant high burnup dry storage cask must be developed for nuclear facilities in Korea to store the increasing high burnup and damaged fuels. More than 20% of fuels generated by PWRs comprise high burnup fuels. This study conducted a structural safety evaluation of the preliminary designs for a high burnup storage cask with 21 spent nuclear fuels and evaluated feasible loading conditions under normal, off-normal, and accident conditions. Two types of metal and concrete storage casks were used in the evaluation. Structural integrity was assessed by comparing load combinations and stress intensity limits under each condition. Evaluation results showed that the storage cask had secured structural integrity as it satisfied the stress intensity limit under normal, off-normal, and accident conditions. These results can be used as baseline data for the detailed design of high burnup storage casks.

A transfer cask serves as the container for transporting and handling canisters loaded with spent nuclear fuels from light water reactors. This study focuses on a cylindrical transfer cask, standing at 5,300 mm with an external diameter of 2,170 mm, featuring impact limiters on the top and bottom sides. The base of the cask body has an openable/closable lid for loading canisters with storage modules. The transfer cask houses a canister containing spent nuclear fuels from lightweight reactors, serving as the confinement boundary while the cask itself lacks the confinement structure. The objective of this study was to conduct a structural analysis evaluation of the transfer cask, currently under development in Korea, ensuring its safety. This evaluation encompasses analyses of loads under normal, off-normal, and accident conditions, adhering to NUREG-2215. Structural integrity was assessed by comparing combined results for each load against stress limits. The results confirm that the transfer cask meets stress limits across normal, off-normal, and accident conditions, establishing its structural safety.

For the sake of future generations, the management of radioactive waste is essential. The disposal of spent nuclear fuel (SNF) is considered an urgent challenge to ensure human safety by storing it until its radioactivity drops to a negligible level. Evaluating the safety of disposal facilities is crucial to guarantee their durability for more than 100,000 years, a period sufficient for SNF radioactivity to become ignored. Past studies have proposed various parameters for forecasting the safety of SNF disposal. Among these, radiochemistry and electrochemistry play pivotal roles in predicting the corrosion-related chemical reactions occurring within the SNF and the structural materials of disposal facilities. Our study considers an extreme scenario where the SNF canister becomes compromised, allowing underground water to infiltrate and contact the SNF. We aim to improve the corrosion mechanism and mass-balance equation compared with what Shoesmith et al. proved under the same circumstances. To enhance the comprehensibility of the chemical reactions occurring within the breached SNF canister, we have organized these reactions into eight categories: mass diffusion, alpha radiolysis, adsorption, hydrate formation, solidification, decomposition, ionization, and oxidation. After categorization, we define how each species interacts with others and calculate the rate of change in species’ concentrations resulting from these reactions. By summing up the concentration change rates of each species due to these reactions, we redefine the mass-balance equations for each species. These newly categorized equations, which have not been explained in detail previously, offer a detailed description of corrosion reactions. This comprehensive understanding allows us to evaluate the safety implications of a compromised SNF canister and the associated disposal facilities by numerically solving the mass-balance equations.

As regulations on carbon emissions increase, the interest in renewable energy is also increasing. However, the efficiency of renewable energy generation is highly low and has limitations in replacing existing energy consumption. In terms of this view, nuclear power generation is highlighted because it has the advantage of not emitting carbon. And accordingly, the amount of spent nuclear fuel is going to increase naturally in the future. Therefore, it will be important to obtain the reliability of containers for transporting safely and storing spent nuclear fuel. In this study, a method for verifying the integrity and airtightness of a metal cask for the safe transportation and storage of spent nuclear fuel was studied. Non-destructive testing, thermal stability, leakage stability, and neutron shielding were demonstrated, and as a result, suitable quality for loading spent nuclear fuel could be obtained. Furthermore, it is meaningful in that it has secured manufacturing technology that can be directly applied to industrial field by verifying actual products.

A radiation shielding resin with thermal stability and high radiation shielding effect has been developed for the neutron shielding resin filled in the shielding shell of dry storage/transport cask for spent nuclear fuel. Among the most commercially available neutron shielding resins, epoxy and aluminum hydroxide boron carbide are used. But in case of the resin, hydrogen content enhances the neutron shielding effect through optimization of aluminum hydroxide, zinc borate, boron carbide, and flame retardant. We developed a radiation shielding material that can increase the boron content and have thermal stability. Flame retardancy was evaluated for thermal stability, and neutron shielding evaluation was conducted in a research reactor to prove the shielding effect. As a result of the UL94 vertical burning test, a grade of V-0 was received. Therefore, it was confirmed that it had flame retardancy. According to an experiment to measure the shielding rate of the resin against neutron rays using NRF (Neutron Radiography Facility), a shielding rate of 91.54% was confirmed for the existing resin composition and a shielding rate of 96.30% for the developed resin composition. A 40 M SANS (40 M Small Angle Neutron Scattering Instrument) neutron shielding rate test was performed. Assuming aging conditions (6 hours, 180 degrees), the shielding rate was analyzed after heating. As a result of the experiment, the developed products with 99.8740% and 99.9644% showed the same or higher performance.

In the event of a loss of a SNF (spent nuclear fuel) transport cask during maritime transportation, it is essential to evaluate the critical depth at which the integrity of the cask can be maintained under high water pressure. SNF transport casks are classified as Type B containers and the integrity of of the containment boundary must be maintained up to a depth of 200 meters unless the containment boundary was breached under beyond-design basis accidents. However, if an intact SNF cask is lost at a depth deeper than 200-meter, release of radioactive material may occur due to breach of containment boundary with over-pressure. In this study, we developed a code for the evaluation of the pressure limit of SNF transport cask, which can be evaluated by inputting the main dimensions and loading conditions of cask. The evaluation model was coded as a computer module for ease of use. In the previous study, models with three different fidelities were developed to ensure the reliability of the calculation and maintain sufficient flexibility to deal with various input conditions. Those three models consisted of a high-fidelity model that provided the most realistic response, a low-fidelity model with parameterized simplified geometry, and a mathematical model based on the shell theory. The maximum stress evaluation of the three models confirmed that the mathematical model provides the most conservative results than the other two models. The previous results demonstrate that mathematical models can be used in the code of computer modules. In this study, additional models of transport cask were created using parametric modeling techniques to improve the accuracy of the pressure limit assessment code for different cask and situations. The same boundary conditions and loading conditions were imposed as in the previous simplified model, and the maximum stress results considering the change in the shape of the transport container were derived and compared with the mathematical model. The comparison results showed that the mathematical model had more conservative values than the simplified model even under various input conditions. Accordingly, we applied the mathematical model to develop a transportation container pressure limit evaluation code that can be simulated in various situations such as shape change and various situations.

Given the domestic situation, all nuclear power plants are located at the seaside, where interim storage sites are also likely to be located and maritime transportation is considered inevitable. Currently, Korea does not have an independently developed maritime transportation risk assessment code, and no research has been conducted to evaluate the release rate of radioactive waste from a submerged transportation cask in the sea. Therefore, secure technology is necessary to assess the impact of immersion accidents and establish a regulatory framework to assess, mitigate, and prevent maritime transportation accidents causing serious radiological consequences. The flow rate through a gap in a containment boundary should be calculated to determine the accurate release rate of radionuclides. The fluid flow through the micro-scale gap can be evaluated by combining the flow inside and outside the transportation cask. In this study, detailed computational fluid dynamic and simplified models are constructed to evaluate the internal flow in a transportation cask and to capture the flow and heat transfer around the transportation cask in the sea, respectively. In the future, fluid flow through the gap will be evaluated by coupling the models developed in this 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.

For safe management of spent nuclear fuels, they should be delivered to repository or waste disposal site. As the amount of spent nuclear fuel transportation is expected to increase in the future due to the provision of an intermediate storage facility, the necessity to secure transportation cask is emerging. In order to secure the spent nuclear fuel transportation cask, it is necessary to analyze the regulatory processes for domestic and foreign spent nuclear fuel transportation cask. In this study, the regulatory processes for domestic and foreign spent nuclear fuel transportation cask was analyzed. In this study, the IAEA, US, and Korea spent nuclear fuel transportation cask regulatory processes were analyzed. The domestic and foreign spent nuclear fuel transportation cask regulatory processes consist of design phase, manufacturing phase, and operation phase. In the design stage, the transport requirements are designed in accordance with the safety requirements of international organizations and countries. The application to be submitted when applying for approval should include a safety analysis report, evidence proving compliance with safety requirements et al. In the manufacturing stage, it is a stage to check whether the safety requirements are satisfied before the first use after manufacturing the transportation cask. Inspections include welding inspection, leakage inspection, shielding inspection, and thermal inspection. In the operation stage, it is a stage of periodically performing inspections for continuous maintenance of the package when the transportation cask is used. The inspection items to be performed are similar to the manufacturing stage and typically include performance inspection of components and leakage inspection. In this study, domestic and foreign spent nuclear fuel transportation cask regulatory processes were analyzed. It was found that the domestic and foreign spent nuclear fuel transportation cask regulatory processes consist of the design phase, the manufacturing phase, and the operation phase. The results of this study can be used as basic data for policy decision-making for the spent nuclear fuel cask.

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.

As the amount of on-site Spent Nuclear Fuel (SNF) in storage increases due to the continued operation of Nuclear Power Plants (NPPs) in Korea, the on-site wet storage pool is expected to become saturated. Therefore, a facility for safely storing the spent nuclear fuel is required so that there is no problem with operation of the NPP until permanent disposal of SNF. Prior to the construction of such a facility, the safety analysis of the interim storage facility and verification of the safety of the spent fuel storage system (e.g. cask, silo) to be used are required according to Article 63 of the Nuclear Safety Act. In this process, analysis of the Structures, Systems, and Components (SSCs) of the storage system is needed. Based on the analysis, it is necessary to efficiently classify SSCs that are important to safety in order to differentiate management that more thoroughly manages those important to safety. In Korea, according to the notice of the Nuclear Safety and Security Commission, the components performing essential safety functions for the safe storage of spent fuel storage system are to be classified as “important safety equipment”. 10 CFR Part 72, a federal regulation related to interim storage facilities in the United States, also requires the identification of SSCs that fall under “Important to Safety (ITS)”, which is like domestic case. In addition, it has been confirmed that there are cases in which detailed classification according to Reg Guide 7.10 and NUREG-CR/6407 is added in Safety Analysis Report. However, these existing classification methods are not only classified as a single grade except for the method according to the Reg guide, but all are classified according to a qualitative standard. Qualitative criteria may cause ambiguity in judgment, resulting in subjective judgment of the person who proceeds in the classification process. Therefore, in this study, a new classification method is proposed to solve the problem according to the qualitative classification method. Assessing the level of radiological harm to the general public due to the assumption of failure of SSC in the spent fuel storage system is used as a quantitative evaluation standard.

Operating and decommissioning nuclear power plants generates radioactive waste. This radioactive waste can be categorized into several different levels, for example, low, intermediate, and high, according to the regulations. Currently, low and intermediate-level waste are stored in conventional 200-liter drums to be disposed. However, in Korea, the disposal of intermediate-level radioactive waste is virtually impossible as there are no available facilities. Furthermore, large-sized intermediate- level radioactive waste, such as reactor internals from decommissioning, need to be segmented into smaller sizes so they can be adequately stored in the conventional drums. This segmentation process requires additional costs and also produces secondary waste. Therefore, this paper suggests repurposing the no-longer-used spent nuclear fuel casks. The casks are larger in size than the conventional drums, thus requiring less segmentation of waste. Furthermore, the safety requirements of the spent nuclear fuel casks are severer than those of the drums. Hence, repurposed spent nuclear fuel casks could better address potential risks such as dropping, submerging, or a fire. In addition, the spent nuclear fuel casks need to be disposed in compliance with the regulations for low level radioactive waste. This cost may be avoided by repurposing the casks.

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.