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.

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.

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.

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.

This presentation summarizes recent research on estimating the mechanical loading environment of spent nuclear fuel (SNF) during normal storage and transportation scenarios sponsored by the US Department of Energy Spent Fuel and Waste Science and Technology (SFWST) program. Normal conditions of truck, ship, and railroad transportation of SNF were studied with testing and numerical modeling to determine that the shock and vibration loads applied to SNF during transportation are not expected to challenge SNF cladding integrity or the fatigue life of cladding. The 30 cm package drop scenario was studied with experiments and modeling to determine that mechanical loads during a 30 cm SNF package drop scenario are only expected to challenge SNF cladding integrity under worstcase conditions at elevated temperatures. The SFWST program is currently preparing seismic shake table testing to record SNF mechanical loads in a dry storage earthquake scenario. This presentation summarizes the findings of the transportation and package drop research and details the progress made on the current seismic test.

The purpose of this study is to develop the analysis procedures for the evaluation of the structural integrity of the spent fuel in normal condition of transport at sea. Spent nuclear fuel must be transported from the wet storage facility in the nuclear power plant to the intermediate storage facility, and the structural integrity must be maintained in vibration and shock loads during the transportation. In general, the transport of spent nuclear fuel is performed in three kinds of modes: road, rail, and sea. During transport, the spent nuclear fuel is subjected to repeated vibration and shock loads by road surfaces, railroad tracks, and waves of the sea. It should be evaluated whether the structural integrity of the spent fuel is maintained under these load conditions. All nuclear power plants in Korea are located in coastal sites, and the interim storage facility for spent nuclear fuel is highly likely to be decided as a coastal site as well. Therefore, the main mode of the spent nuclear fuel transport is expected to be maritime transport by ships. In this study, the analysis procedure was developed to evaluate the safety of spent fuel at maritime transport by ships, and the procedure for evaluating the integrity of spent fuel under normal conditions of maritime transport were proposed. CFD analysis using SeaFEM was performed for the vibration analysis of the ship by waves, and the structural vibration analysis of the transport system was simulated using the developed in-house codes. The fatigue durability of the cladding was also evaluated using the developed fatigue analysis program and the fatigue analysis used the strain data obtained from the structural analysis. It was concluded that the value of the fatigue damage on the spent fuel cladding during normal conditions of maritime transportation is close to “0” and the structural integrity of the spent fuel is maintained in the same condition.

Sandia National Laboratories is the lead laboratory for the United States Department of Energy for the research and development (R&D) efforts to support the technical basis for the long-term storage, subsequent transportation, and permanent disposal of commercial spent nuclear fuel and high-level waste. Sandia does not design nuclear facilities; Sandia performs R&D to help ensure facilities and the fuel cycle are safe, sustainable, and secure. This talk will focus on the spent fuel storage and transportation programs that contribute to this work. The goal in spent fuel storage and transportation R&D is to understand the mechanical integrity of the fuel, cladding, and storage system beyond interim storage and into disposal time frames. Our research is focused on understanding the high burn-up cladding integrity over time, understanding the thermal behavior during drying and storage, understanding potential cladding oxidation pathways, and quantifying in the external loads experienced during transportation, handling, and seismic events. Additionally, this work includes extensive work to understand the basic science of canister stress corrosion cracking and the potential consequences of a through wall canister crack.

This paper aims to evaluate the mechanical integrity for Spent Nuclear Fuel (SNF) cladding under lateral loads during transportation. The evaluation process requires a conservative consideration of the degradation conditions of SNF cladding, especially the hydride effect, which reduces the ductility of the cladding. The dynamic forces occurring during the drop event are pinch force, axial force and bending moment. Among those forces, axial force and bending moment can induce transverse tearing of cladding. Our assessment of 14 × 14 PWR SNF was performed using finite element analysis considering SNF characteristics. We also considered the probabilistic procedures with a Monte Carlo method and a reliability evaluation. The evaluation results revealed that there was no probability of damage under normal conditions, and that under accident conditions the probability was small for transverse failure mode.

This study developed an analytical methodology for the mechanical integrity of spent nuclear fuel (SNF) cladding tubes under external pinch loads during transportation, with reference to the failure mode specified in the relevant guidelines. Special consideration was given to the degraded characteristics of SNF during dry storage, including oxide and hydride contents and orientations. The developed framework reflected a composite cladding model of elastic and plastic analysis approaches and correlation equations related to the mechanical parameters. The established models were employed for modeling the finite elements by coding their physical behaviors. A mechanical integrity evaluation of 14 × 14 PWR SNF was performed using this system. To ensure that the damage criteria met the applicable legal requirements, stress-strain analysis results were separated into elastic and plastic regions with the concept of strain energy, considering both normal and hypothetical accident conditions. Probabilistic procedures using Monte Carlo simulations and reliability evaluations were included. The evaluation results showed no probability of damage under the normal conditions, whereas there were small but considerably low probabilities under accident conditions. These results indicate that the proposed approach is a reliable predictor of SNF mechanical integrity.