A comparison and validation between the analysis and vibration test data of a nuclear fuel assembly were conducted. During the comparison and validation process, various parameters that govern the vibration behavior of the fuel assembly were determined, including nuclear fuel rod’s stiffness, spring constants of the dimple and spring of support structures, and damping coefficients. The calibration of the vibration analysis model aimed to find analysis parameters that can accurately simulate the vibration behavior of the test data. For calibration, power spectral density (PSD) diagrams were generated for both the measured signals from the test and the calculated signals from the analysis. The correlation coefficient between these two PSD plots was calculated. To find the analysis parameters, each parameter was defined as a variable with an appropriate range. Latin hypercube sampling was used to generate multiple sample points in the variable space. Analysis was performed for the generated sample points, and PSD plot correlation coefficients were calculated. Using the generated sample points and their corresponding results, a Gaussian Process Regression model was implemented for PSD plot correlation coefficients and the maximum PSD value. Based on the constructed surrogate model, the optimal analysis parameters were easily found without additional computations. Through this method, it was confirmed that the analysis model using the optimal parametes appropriately simulates the vibration behavior of the test.

In this study, a fracture evaluation of the spent nuclear fuel storage canister was conducted. Stainless steel alloys are typically used as the material for canisters, and therefore, a separate destructive evaluation is not required for safety analysis reports. However, in this research, a methodology for conducting a destructive evaluation was proposed for assessing the acceptability of cracks detected during in-service inspections for long-term storage due to reasons such as stress corrosion cracking. For the fracture evaluation, analytical equations provided in the design code such ASME were employed, and finite element method (FEM) based linear elastic fracture mechanics (LEFM) was performed to validate the effectiveness of the analytical equations. Impact analyses such as tip-over of the storage cask on a concrete pad were performed, and the fracture evaluation using stresses resulting from the impact analysis under accident conditions and residual stresses from welds were carried out. Through this research, geometric dimensions for cracks exceeding the fracture criteria were established.

Concrete structures of spent nuclear fuel interim storage facility should maintain their ability to shield and structural integrity during normal, off-normal and accident conditions. The concrete structures may deteriorate if the interim storage facility operates for more than several decades. Even if deterioration occurs, the concrete structures must maintain their own functions such as radiation shielding protection and structural integrity. Therefore, it is necessary to establish an analysis methodology that can evaluate whether the deteriorated concrete structure maintains its integrity under not only normal or off-normal condition but also accident condition. In this study, dynamic material testing was conducted on concrete cores extracted from HANARO exterior wall during seismic reinforcement construction. HANARO was constructed at the Korea Atomic Energy Research Institute in 1995, following strict nuclear quality assurance standards. In order to conduct the dynamic material testing of the extracted concrete cores, self-disposal had to be performed because the concrete cores were extracted and stored in a radiation controlled area. A self-disposal application was prepared and submitted based on the radionuclide analysis results, and it was finally approved in April 2023. Then, a test was performed by processing a specimen for dynamic property testing using a self-disposed concrete core. The concrete cores were processed to create specimens for dynamic material testing and the dynamic material testing was performed to obtain stress-strain diagrams according to the strain rate.

This paper described a method for analyzing the structural performance of a metal container used for disposing radioactive waste generated during the decommissioning of a nuclear power plant, and numerical analysis results of a method for reinforcing the container. The containers to be analyzed were those that can be used in near-surface and landfill disposal facilities scheduled to be operated at the Gyeongju radioactive waste disposal facility. Structural reinforcement of the container was performed by lattice reinforcement, column reinforcement, and bottom plate reinforcement. Accordingly, a total of 14 reinforcement cases were modeled. The external force causing damage to the container was set equivalent to the impact of a 9-m fall, accounting for the height of the vault at the near-surface disposal facility. The reinforcement methods with a high contribution to the structural performance of the container were concluded to be lattice and column reinforcements.

Transport packages have been developed to transport the decommissioning waste from the nuclear power plant. The packages are classified with Type IP-2 package. The IAEA requirements for Type IP-2 packages include that a free drop test should be performed for normal conditions of transport. In this study, drop tests of the packages were performed to prove the structural integrity and to verify the reliability of the analysis results by comparing the test and analysis results. Half-scale models were used for the drop tests and drop position was considered as 0.3 m oblique drop on packages weighing more than 15 tons. The strain and impact acceleration data were obtained to verify the reliability of the analysis results. Before and after the drop tests, radiation shielding tests were performed to confirm that the dose rate increase was within 20% at the external surface of the package. Also, measurement of bolt torque, and visual inspection were performed to confirm the loss or dispersion of the radioactive contents. After each drop test, slight deformations occurred in some packages. However, there was no loss of pretension in the lid bolts and the shielding thickness was not reduced for metal shields. In the package with concrete shield, the surface dose rate did not increase and there was no cracks or damage to the concrete. Therefore, the transport packages met the legal requirements (no more than a 20% increase of radiation level and no loss or dispersion of radioactive contents). Safety verifications were performed using the measured strain and acceleration data from the test, and the appropriate conservatism for the analysis results and the validity of the analysis model were confirmed. Therefore, it was found that the structural integrity of the packages was maintained under the drop test conditions. The results of this study were used as design data of the transport packages, and the packages will be used in the NPP decommissioning project in the future.

Waste containers for packaging, transportation and disposal of NPP (Nuclear Power Plant) decommissioning wastes are being developed. In this study, drop tests were conducted to prove the safety of containers for packaging of the wastes and to verify the reliability of the analysis results by comparing the test and analysis results. The drop height of the waste containers was considered to be 30 mm, which is the maximum lifting speed of a 50 tons crane in the waste treatment facility converted to the drop height. Drop orientation of the containers was considered for bottom-end on drop. The impact acceleration and strain data were obtained to verify the reliability of the analysis results. Before and after the drop tests, measurement of the dose rate and the radiographic testing for concrete wall, and measurement of the wall thickness of steel plate were conducted to evaluate the radiation shielding integrity. Also, measurement of bolt torque, and visual inspection were conducted to evaluate the loss or dispersion of radioactive contents. After the drop tests, the radiation dose rate on the container surface did not increase by more than 20%, and there was no crack in the concrete. In addition, the thickness of the steel plate did not change within the measurement error. Therefore, the radiation shielding integrity of the container was maintained. After the drop tests, the lid bolts were not damaged and there was no loss of pretension in the lid bolts. In addition, there was no loss or dispersion of the contents as a result of visual inspection. In order to prove the reliability of the drop analysis results, safety verifications were performed using the drop test results, and the appropriate conservatism for the analysis results and the validity of the analysis model were confirmed. Therefore, the structural integrity of the waste containers was maintained under the drop test conditions.

The design of a wooden impact limiter equipped to a transportation cask for radioactive materials was optimized. According to International Atomic Energy Agency Safety Standards, 9 m drop tests should be performed on the transportation cask to evaluate its structural integrity in a hypothetical accident condition. For impact resistance, the size of the impact limiter should be properly determined for the impact limiter to absorb the impact energy and reduce the impact force. Therefore, the design parameters of the impact limiter were optimized to obtain a feasible optimal design. The design feasibility criteria were investigated, and several objectives were defined to obtain various design solutions. Furthermore, a probabilistic approach was introduced considering the uncertainties included in an engineering system. The uncertainty of material properties was assumed to be a random variable, and the probabilistic feasibility, based on the stochastic approach, was evaluated using reliability. Monte Carlo simulation was used to calculate the reliability to ensure a proper safety margin under the influence of uncertainties. The proposed methodology can provide a useful approach for the preliminary design of the impact limiter prior to the detailed design stage.

In this study, a drop analysis of metallic disposal containers for radioactive wastes is performed according to accident scenarios at the disposal site. The weight of the disposal container is about 8 tons, and the ingot-type wastes are loaded in the disposal container. To simulate the floor of the disposal site as the impact target, the reinforced concrete pad is modeled. High impact energy of the disposal container due to their heavy weight and high drop height causes excessive deformation and failure of the concrete target having relatively weak strength. Dynamic growth of cracks due to such failures causes penetration and delamination of concrete. Since the impact force delivered to the container strongly depends on the failure of the concrete pad, it is important to properly simulate the failure of the concrete in the drop analysis. A material erosion method can be used to simulate the concrete failure. In the case of applying erosion based on the finite element method (FEM), the element is deleted when the element exceeds a certain criterion, which causes material and energy loss problem. To solve this problem, mesh-free methods such as smoothed particle hydrodynamics (SPH) can be commonly used, but the mesh-free method has the disadvantage of incurring high numerical cost. Therefore, an adaptive method combining SPH and FEM-based SOLID elements is used for concrete target modeling to simulate excessive deformation and failure of the concrete target. In the adaptive coupling method of SPH and SOLID, the concrete target is first modeled as a solid element. When the damage of concrete exceeds the failure criterion, the solid element is eroded and the SPH element replacing the solid element is activated. Since the activated SPH element continues to participate in the impact, the problem of loss of materials and energy can be effectively solved. In this way, analysis results consistent with actual physical phenomena can be obtained.

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.