Since spent nuclear fuel (SNF) should be isolated from the human life zone for at least 106 years, deep geological disposal (DGD) is considered a strong candidate for SNF management in many countries. Therefore, a disposal canister should be nearly immune to corrosion in such a long-term storage environment. Even though copper has a low corrosion rate of a few millimeters per million years in geological environments, the corrosion resistance of the copper welds must be preferentially validated, which inevitably occurs during the sealing of the disposal canister after the SNF is loaded. This is because the weld zone is a discontinuous area of microstructure, which can accelerate uniform and localized corrosion. In this study, the microstructural characteristics of copper welds in different welding conditions such as friction stir welding, electron beam welding, cold spray, were analyzed, focusing on the formation of microstructure, which affects resistance to corrosion. In addition, the microstructure and corrosion properties of the copper weld zone manufactured by recent wire-based additive manufacturing (AM) technology were experimentally evaluated. From this preliminary test result, it was found that the corrosion characteristics of the welds produced by the AM process using wire are comparable to those of the conventional forged copper plate.

Geologic disposal at deep depth is an acceptable way to dispose of high-level radioactive waste and isolate it from the biosphere. The geological repository system comprises an engineered barrier system (EBS) and the host rock. The system aims to delay radionuclide migration through groundwater flow, and also, the flow affects the saturation of the bentonite in the EBS. The thermal conductivity of bentonite is a function of saturation, so the temperature in the EBS is directly related to the flow system. High-temperature results in the two-phase flow, and the two-phase flow system also affects the flow system. Therefore, comprehending the influencing parameters on the flow system is critical to ensure the safety of the disposal system. Various studies have been performed to figure out the complex two-phase flow characteristics, and numerical simulation is considered an effective way to predict the coupled behavior. DECOVALEX (DEvelopment of COupled models and their VALidation against EXperiments) is one of the most famous international cooperating projects to develop numerical methods for thermo-hydro-mechanicalchemical interaction, and Task C in the DECOVALEX-2023 has the purpose of simulating the Fullscale Emplacement (FE) experiment at the Mont-Terri underground research laboratory. We used OGS-FLAC, a self-developed numerical simulator combining OpenGeoSys and FLAC3D, for the simulation and targeted to analyze the effecting parameters on the two-phase flow system. We focused on the parameters of bentonite, a key component of the disposal system, and analyzed the effect of compressibility and air entry pressure on the flow system. Compressibility is a parameter included in the storage term, defining the fluid storage capacity of the medium. While air entry pressure is a crucial value of the water retention curve, defining the relation between saturation and capillary pressure. From a series of sensitivity analyses, low compressibility resulted in faster flow due to low storage term, while low air entry pressure slowed flow inflow into the bentonite. Low air entry pressure means the air easily enters the medium; hence the flow rate becomes lower based on the relativity permeability definition. Based on the sensitivity analysis, we further investigate the effect of shotcrete around the tunnel and excavation damaged zone. Also, long-term analysis considering heat decay of the radioactive waste will be considered in future studies.

Backfill is one of the key elements of deep geological disposal. The backfill material is used to fill disposal tunnels and is mainly composed of swellable clay, preventing the migration of nuclide and structurally supporting the tunnel. The selection and application of backfill material are critical for the stable and efficient disposal of spent fuel. Therefore, it is essential to secure various candidate materials for backfill and to comprehensively understand the properties and behavior of these materials. Recently, the Korea Atomic Energy Research Institute has selected a candidate material called Bentonil-WRK and is evaluating its applicability. To utilize this material as backfill, the safety function of a mixed backfill concept, consisting of sand and Bentonil-WRK, was assessed. The swelling pressure was measured as a function of dry density for a bentonite/silica sand mix ratio of 3/7. The results showed that the swelling pressure ranged from 0.15 to 0.273 MPa, depending on the dry density, with higher dry densities resulting in higher swelling pressures. The measured swelling pressure met the target performance criteria suggested by SKB and Posiva (i. e., 0.1 MPa), but did not meet the design requirement for swelling pressure (i. e., 1 MPa). This indicate the need for further research after increasing the mass fraction of bentonite (e. g., mix ratio 4/6 or more). The results of this study are expected to be used in the selection of candidate backfill materials and the establishment of design guidelines for engineered barrier backfill.

A methodology is under development to reconstruct and predict the long-term evolution of the natural barrier comprising the site of radioactive waste disposal. The natural barrier must protect the human zone from radionuclides for a long time. So for this, we need to be able to restore the evolution of the bedrock constituting the natural barrier from the past to the present and to predict from the present to the future. A methodology is being studied using surface outcrop, tunnel face of KURT (KAERI Underground Research Tunnel), and drill core at KAERI (Korea Atomic Energy Research Institute). Among them, drill core is an essential material for identifying deep geological properties, which could not be confirmed near the surface when considering the geological condition of the repository in the deep part. In this study, we selected several qualitative and quantitative analyses to construct a deep lithological model from the disposal perspective. These were applied to drill core samples around the KURT. There are the dikes presumed the Cretaceous were intruded by Jurassic granitoids in the study area. Analyzing trace elements of each rock type in the study area classified through geochemical characteristics and microstructure in previous studies made it possible to obtain qualitative information on the petrogenetic process. In addition, synthesizing the quantitative numerical age allows for grasping the evolution of bedrock, including intrusion and cutting relationships. LAICPMS was used for determining the age of zircons in plutonic rocks. The highly reliable 40Ar-39Ar method was selected for volcanic rocks because it can correct the loss of Ar gas and obtain the values of two types of Ar isotopes in a single sample. As a result, it was possible to infer the formation environment of rocks through anomalies in specific trace element content. And according to the numerical ages, it was possible to support the known separated rock type found in previous studies or to present a quantitative precedence relation for unclassified rocks. These methods could be applied to reconstruct the long-term evolution of bedrock within natural barriers.

In the engineered barrier system of deep geological disposal repository, complex physicochemical phenomena occur throughout the entire disposal time, consequently impacting the safety function. The bentonite buffer, a significant component of the engineered barrier system, can be geochemically altered due to the changes in host rock groundwater, temperature, and redox condition. Such changes may have direct or indirect effects on radionuclide migration in case of canister failure. Therefore, a modeling tool that accounts for coupled thermal-hydraulic-mechanical-chemical (THMC) processes is necessary for the safety assessment. To this end, the Korea Atomic Energy Research Institute (KAERI) has developed the APro, a modeling interface for conducting safety assessment of deep geological disposal repository. The APro considers coupled THMC processes that influence radionuclide migration. Here, the solute transport considering thermal and hydraulic processes are calculated using the COMSOL multi-physics, while geochemical reactions are carried out in PHREEQC. The two software are coupled using a sequential non-iterative operator splitting approach, and transport of non-water H, non-water O, and charge were additionally considered to enhance the coupling model stability. Finally, the applicability of APro to simulate long-term geochemical evolution of bentonite was demonstrated through benchmark studies to evaluate the effects of mineral precipitation/dissolution, temperature, redox, and seawater intrusion.

The purpose of this study was to examine whether galvanic corrosion of copper occurs by inserting a third barrier layer with a higher corrosion potential than copper between copper and cast iron when the copper layer is locally perforated by pitting or partial corrosion. A triple layer composed of copper, inserted metal, and carbon steel was manufactured by cold spray coating of inserting metal powders such as Ag, Ni, and Ti on carbon steel plate followed by Cu coating on it. First, the corrosion properties were evaluated electrochemically for each metal coating. As a result of Tafel plot anaylsis in KURT groundwater condition, the corrosion potential of Fe (-567 mV) was much lower than that of Cu (-91 mV), and the corrosion potential of Ni (-150 mV) was also lower than that of Cu. Therefore, Ni was likely to corrode before Cu. However, the corrosion current of Ni was lower than that of the Cu. In the galvanic specimen where the copper and inserting metal were exposed together, Cu-Fe was much lower corrosion potential of -446 mV, and the corrosion potential of Cu-Ti, Cu-Ni, and Cu-Ag were slightly higher than that of Cu. Therefore, it seemed that Ag, Ni, and Ti all might promote galvanic corrosion of surrounding copper when the copper layer was perforated to the inserted metal layer. If the metal insertion presented in this study operates properly, the disposal container does not need to worry about the partial corrosion or non-uniform corrosion of external copper layer.

IAEA safety standards document and international programs (such as BIOMASS) related to the assessment of the biosphere around High Level Radioactive Waste (including Spent Nuclear Fuel) repositories require the assessment of the biosphere to use the assumption that the current natural environment and human society will be maintained, and at the same time, the evolution of the distant future changes also need to be taken into account. In Korea, which has not designated candidate disposal sites, it is necessary to investigate and predict the current state and future changes of the natural environment throughout Korea and apply it practically to Biosphere assessment (for BDCF derivation) for candidate disposal sites suitability assessment and Safety Case (for performance assessment) preparation for design, construction, operation, and post-closure management. To this end, the natural environment in the fields of Topography, Geology, Soil, Ecology, Weather and Climate, Animals and Plants, Hydrology, Ocean, Land-use, etc. and human society in the fields of Population Distribution, Spatial-Planning, Urban Form, Industrial-Structure, Lifestyle etc. are being investigated in the context of current status, past change records, and future change potential in the Korean Peninsula. This paper summarizes those investigations to date. This study referred Biomass-6 [IAEA] and National Atlas I (2019)/II (2020)/III (2021) [National Geographic Information Institute of the Korea Ministry of Land, Infrastructure and Transport].

To obtain a license for a deep geological disposal repository for spent nuclear fuel, it is necessary to perform a safety assessment that quantifies the radiological impact on the environment and humans. One of the key steps in the safety assessment of a deep geological repository is the development of scenarios that describe how the repository evolves over the performance period and how events and processes affect performance. In the field of scenario development, demonstrating comprehensiveness is critical, which describes whether all factors that are expected to have a significant impact on the repository's performance have been considered. Mathematical proof of this is impossible. However, If the scenario development process is logical and systematic, it can support the claim that the scenario is comprehensive. Three primary approaches are being considered for scenario development: ‘Bottomup’, ‘Top-down’, and ‘Hybrid’. Hybrid approach provides a more systematic and structured process by considering both the FEPs (Features, Events, Processes) and safety functions utilized in the bottomup and top-down approaches. Many countries that develop recent scenarios prefer demonstrating scenario comprehensiveness using a hybrid approach. In this study, a systematic and structured scenario development process of a hybrid approach was formulated. Based on this, sub-scenarios were extracted that describe the phenomena occurring in the repository over the performance period, categorized by period. By integrating and screening the extracted sub-scenarios, a scenario describing the phenomena occurring over the entire period of disposal was developed.

In Korea, borated stainless steel (BSS) is used as a storage rack in spent fuel pools (SFP) to maintain the nuclear criticality of spent fuels. As the number of nuclear power plants and the corresponding amount of spent fuels increased, the density in SFP storage rack also increased. In this regard, maintaining subcriticality of spent nuclear fuels became an issue and BSS was selected as the structural material and neutron absorber for high density storage rack. Since it is difficult to replace the storage rack, corrosion resistance and neutron absorbency are required for long period. BSS is based on stainless steel 304 and is specified in the ASTM A887-89 standard depending on the boron concentration from 304B (0.20-0.29% B) to 304B7 (1.75-2.25% B). Due to the low solubility of boron in austenitic stainless steel, metallic borides such as (Fe, Cr) 2B are formed as a secondary phase. Metallic borides could cause Cr depletion near it, which could decrease the corrosion resistance of the material. In this paper, the long-term corrosion behavior of BSS and its oxide microstructures are investigated through accelerated corrosion experiment in simulated SFP conditions. Because the corrosion rate of austenitic stainless steel is known to be dependent on the Arrhenius equation, a function of temperature, the corrosion experiment is conducted by increasing the experimental temperature. Detail microstructural analysis is conducted using a scanning electron microscope, transmission electron microscope and energy dispersive spectrometer. After oxidation, a hematite structure oxide film is formed, and pitting corrosion occurs on the surface of specimens. Most of the pitting corrosion is found at the substrate surface because the corrosion resistance of the substrate, which has low Cr content, is relatively low. Also, the oxidation reaction of B in the secondary phase has the lowest Gibbs free energy compared to other elements. Furthermore, oxidation of Cr has low Gibbs free energy, which means that oxidation of B and Cr could be faster than other elements. Thus, the long-term corrosion might affect the boron content and the neutron absorption ability of the material. Using boron’s high cross-section for neutrons, the neutron absorption performance of BSS was evaluated through neutron transmission tests. The effect of the corrosion behavior of BSS on its neutron absorption performance was investigated. Samples simulated to undergo up to 60 years of degradation before corrosion through accelerated corrosion testing did not show significant changes in the neutron shielding ability before and after corrosion. This can be explained in relation to the corrosion behavior of BSS. Boron was only leached out from the secondary phase exposed on the surface, and this oxidized secondary phase corresponds to about 0.17% of the volume of the total secondary phase. This can be seen as a very small proportion compared to the total boron content and is not expected to have a significant impact on neutron absorption performance.

South Korea has been storing UNF in spent fuel pool dry storage facility within Nuclear Power Plants. The dry storage facility of used nuclear fuel (UNF) is essential to sustain safety and sustain stable operation of a nuclear power plant. Most abroad countries have attempted to develop a variety of dry storage facility for used nuclear fuel in order to retain the safe restoration. Many studies have been conducting to safety evaluation for the dry storage facility. However, there is not a ventilation evaluation in the wake of fire event that could influence of the thermal effect on the dry storage facility, even though it will likely to occur fire events such as wildfire, air craft crash. In practice, it happened to catastrophic disaster due to the wild fire adjacent to ul-jin mountain. Also, it happened to fire accident near to the Japonia NPP in Ukraine territory caused of military air plane missile. It has not mostly been studied on the ventilation evaluation considered to thermal safety in the dry storage facility excepted for some researches. It could need the mechanical ventilation systems such as HVAC system in the dry storage system, so that thermal effect can be reduced. In this study, we conducted to the ventilation control modelling by using fire modelling tool (Fire Dynamic Simulator v.6.7). The ventilation scenarios made up for 3 case that can compare flowrate variation with ventilation control. As a result of modelling, there is no differentiation between ventilation control using performance curve with not using performance curve even though the pressure fluctuation would be increased, compared with the case of considering performance curve. Second, it evaluated that the mode for fraction control would occur to pressure rise in the state of controlling the ventilation system flowrate. However, sensitivity of flowrate control was more decreased below less than 5 seconds. Third, in the case of on/off control system revealed more higher resolution than other cases caused by flowrate variation. These results could be considered as the design guidelines for the development dry storage facility to improve the thermal performance that can reduce thermal risk. Furthermore, the study results would expect HVAC system installed in dry storage to help automatic ventilation control relevant to dry storage safety increased.

Korea Atomic Energy Research Institute (KAERI) has been operating the Post Irradiation Examination Facility (PIEF) for spent fuel. The facility has pools and hot cells for handling and examining fuel assemblies and rods. In the first hot cell, non-destructive tests such as visual inspection, defect detection, oxide layer thickness measurement, and gamma scanning are performed on a full-length fuel rod. Then, the fuel rod is transported to the next hot cell for measuring the rod internal pressure (RIP). After the RIP measurement, the fuel rod is cut by a cutting machine to make samples for destructive tests. Currently, the existing cutting machine is broken, so a new machine needed to be designed and manufactured. The major considerations for designing the cutting machine were convenience of remote handling and decontamination. The machine was modularized and its handling parts were designed to be easily controlled by manipulators. The cover was designed to prevent radioactive contamination of the surrounding area.

Korean MMTT project has been launched in order to clarify the vibration and shock loads under normal condition and transportation (NCT) in Korean geological and transportation conditions and to evaluate the integrity of SNF under such a transportation load. To evaluate the integrity of the SNF during normal land and sea transport tests, a representative SNF that represents the entirety of the different types of SNFs stored in the spent fuel pool of the power plant should be selected. And, it is necessary to make the test assembly to have a statically and dynamically similar behavior with the actual SNF. Therefore, in this project, we selected two types of fuel assembly that are expected to exhibit relatively conservative behavior under NCT, and these assemblies are being fabricated into surrogate test assemblies to have a similar characteristic as actual SNF based on the accumulated data from the poolside examination and the hot cell test so far. Tests were conducted for NCT conditions. In addition, a fatigue test was performed to integrity of the nuclear fuel rods under NCT conditions. Nuclear fuel assemblies are transported while being laid inside the cask under NCT, and are exposed to external shocks and vibrations. At this time, the fuel rod between the grid and grid is exposed to bending motion by this external force. For this simulation, a fixture was developed and used for static bending tests and bending fatigue tests. To simulate spent nuclear fuel rod specimens, hydrogen reorientation Zry-4 cladding was used and simulated pellets made of stainless steel were applied. And also, it was bonded using epoxy to give bonding conditions between the inside and the pellet. As a result of the test, cracks occurred due to the concentrated load between the pellets, resulting in damage to the fuel rod. The fatigue results showed a similar trend compared to the results performed by ORNL, and the lower bound fatigue curve presented by NUREG-2224 was also satisfactory.

In this study, radioactivity of Cs-134, Cs-137, and Eu-154, which are gamma-emitting nuclides among fission products of spent fuel, was analyzed as a tool to measure the burnup of spent fuel nondestructively. This nuclide has a unique gamma-ray energy, allowing the amount of the isotope to be estimated based on the intensity of the gamma-ray at a specific energy. The SCALE 6.2 ORIGAMI (ORIGen AsseMbly Isotopics) module and the latest ORIGEN-arp library were used for computational analysis. The spent fuel samples were selected as WH14×14 with an enrichment of 1.5~5.0wt%, a burnup of 10~60 GWD/MTU, and a cooling time of 0~40 years. The analysis results were benchmarked using SFCOMPO experimental data provided by OECD/ NEA, including isotope inventory and uncertainty measured by destructive radiochemical analysis, fuel assembly design data required for benchmark model development, reactor design information, and operating history information. 16 similar spent fuels were selected from SFCOMPO data and the calculation results of Cs-134, Cs-137, and Eu-154 were compared. The average error of the Cs-134 radioactivity calculation result was 2.81%, and the maximum error was 6.70%. The average errors of Cs-137 and Eu-154 were 2.42% and 4.95%, respectively, and the maximum errors were 5.40% and 14.91%, respectively.

Spent Fuel Pool Island (SFPI) is a spent nuclear fuel storage pool that operates independently of existing nuclear facilities to safely manage SNF and minimize maintenance costs during the nuclear decommissioning process. Since the radiation controlled area can be dismantled before transporting SNF to a dry storage facility, the overall decommissioning period can be shortened, and the risk of occupational exposure during dismantling is reduced. In the US, various nuclear power plants have introduced SFPI for this reason. In this paper, to analyze the economic feasibility of application of SFPI to nuclear power plants to be decommissioned, several scenarios are established in consideration of the decommissioning plan and schedule, SFPI and dry storage facility application schedule. Cost and benefit list (SFPI application cost, SNF management cost, SNF dry storage cask cost, etc.) for each alternative were derived, and economic analysis was conducted by applying the Net Present Value (NPV). As a result of the analysis, it is found that the application of SFPI during decommissioning is economically effective as the NPV showed a positive number even when uncertainty was taken into account.

In case of damaged spent fuels, it would require additional treatment for their transportation and storage to capture the radioactive fission products in a defined space. The canning container for the damaged spent fuels is one way to seal the radioactive fission products inside the container. In the Post Irradiation Examination Facility (PIEF) of KAERI, the Quiver container has been introduced for canning damaged spent fuels from Westinghouse Sweden. The main container body has been manufactured for particle-tightness of spent fuel. In addition, drying equipment is being prepared for gas-tightness of spent fuel. The drying equipment can remove water and fill the inert gas inside the container. Before drying inside the container, we evaluated the volatile fission products inventory because volatile fission products could be released during the drying process. Despite assuming highly conservative hypotheses for the inventory remaining in damaged fuel rods, the amount that could be released during the drying process was less and dose rate levels around the evacuation piping system were low.

The current storage capacity of the spent nuclear fuel at the Kori unit 2 has reached approximately 94% saturation, excluding emergency core capacity. As South Korea has not yet constructed any interim storage facilities to store spent nuclear fuel, one of possible options is to install high density storage racks in spent fuel pool at the reactor site to increase its capacity. The high density storage rack is a technology developed to allow the storage to have more spent nuclear fuel than originally planned, while still ensuring safety. It achieves this by reducing the storage gap between the spent nuclear fuel. The purpose of this study is to investigate an appropriate storage capacity for spent fuel pool in the Kori unit 2 and the factors to be considered during the high density storage rack installation. Given that the Kori unit 2 is planning continued operation and Korea Hydro & Nuclear Power (KHNP) is preparing to install a temporary storage facility for spent nuclear fuel at the Kori nuclear site, it is reasonable to consider the installation of high density storage racks in the spent fuel pool as a storage solution for these issues. When evaluating the capacity of the spent fuel pool, the amount of spent nuclear fuel generated by other reactors in Kori nuclear site and the amount of spent nuclear fuel generated by continued operation of the Kori unit 2 should be taken into account. This study aims to consider these factors and evaluate the capacity of the spent fuel pool. Furthermore, when installing high density storage rack for the spent nuclear fuel, it should be noted that the existing storage racks at the Kori unit 2 are welded to the liner plate, which may require additional cutting work. Therefore, it is necessary to review the suitable method for the cutting work. Additionally, assuming that divers need to access the area near the storage racks or cutting & welding devices require radiation protection in the area, it is essential to analyze the expected radiation level with computational code and propose appropriate measures to limit work time or establish a work zone. Thus, this study evaluates appropriate capacity of spent fuel pool and work methods for the installation of high density storage rack in the spent fuel pool at the Kori unit 2. It is expected that this paper contributes to install high density storage racks in SFP safely.

CANDU Spent Fuel (CSF) dry storage system, SILO, has been operated from 1992 at Wolsung under 50 year operating license. As of 2023, this system has been operated for over 30 years and its licensed remaining operation time is less than 20 years. When it faces the final stage of operation, it has only two options; moving to a centralized away-from-reactor storage or extending its license atreactor. These two options have an inevitable common duty of confirming the CSF integrity by a “demonstration test”. Since the degradation of CSF and structural materials in the SILO are critically dependent on temperature, two important goals of the ‘DEMO test’ were set as follows. 1. Design of ‘DEMO SILO’: Development of internal monitoring technology by transforming SILO design. 2. Accurate measurement and evaluation of the three-dimensional temperature distribution in the ‘DEMO SILO’ Based on operating real commercial SILO dimension, a conceptual “DEMO SILO” design has been developed from 2022. Because, unlike with commercial Silo, ‘Demo Silo’ must be disassembled and assembled, and have penetration holes. Safety evaluation technologies like structural, thermal and radiation protection analysis also have been developed with design work. ‘Demo SILO’ should evaluate an accurate 3D temperature distribution with minimal number of thermocouples and penetration holes to avoid disruption of internal flow and temperature distribution. For this reason, a ‘Best Estimate Thermal-Hydraulics evaluation system for SILO’ is under development and it will be essential for ensuring temperature prediction accuracy. Construction of a full-scale test apparatus to validate this technology will begin in 2024. In order to supply power to many heaters and monitor temperature gradient inside of this apparatus, it has modular design concept by dividing its whole body to axial 9 sub-bodies which looks like a donut containing a basket at center position.

Flow-induced vibration can lead to fretting wear damage of fuel rods and spacer grids in nuclear reactors. Similarly, during the transport of spent nuclear fuel assemblies, continuous vibration and intermittent impact might also result in fretting wear due to dynamic interaction. Therefore, it is important to evaluate fuel rod damage due to fretting wear under such transport conditions. This study examines spent nuclear fuel rod specimens fabricated with hydride cladding tubes and simulated pellets, with hydrogen content ranging from 200 to 700 ppm and oxide film thickness ranging from 25 to 100 micrometers. Tests were conducted under a worst-case scenario, assuming continuous exposure to that condition during the expected transport time. Results indicate that the wear depth of all rod specimens occurred within the oxide film, suggesting a high resistance to fretting wear during transportation.

There have been a variety of issues related to spent nuclear fuel in Korea recently. Most of the issues are related to intermediate storage and disposal of spent nuclear fuel. However, recently, various studies have been started in advanced nuclear countries such as the United States to reduce spent nuclear fuel, focusing on measures to reduce spent nuclear fuel. In this study, a simple preliminary assessment of the thermal part was performed for the consolidation storage method which separates fuel rods from spent nuclear fuel and stores them. The preliminary thermal evaluation was analyzed separately for storing the spent fuel in fuel assembly state and separating the fuel rods and storing them. The consolidation storage method in separating the fuel rods was advantageous in terms of thermal conductivity. However, detailed evaluation should be performed considering heat transfer by convection and vessel shape when storing multiple fuel bundles simultaneously.

After spent fuel is stored in a dry storage container, it becomes difficult to obtain information on the fuel’s characteristics. As a result, it is necessary to identify the characteristics of spent nuclear fuel in advance and secure the information necessary to establish delivery acceptance requirements for interim storage and disposal in the future. Therefore, it is necessary to evaluate the characteristics of spent fuel before loading dry storage casks. In order to prepare for the dry storage of spent fuel, information on the basic characteristics of the fuel is required. As part of this information, it is also necessary to establish calculation criteria for spent fuel burnup. Spent fuel burnup can be classified into three categories. The first is burnup evaluated using design codes (design burnup), the second is burnup measured by furnace instruments during power plant operation (actual burnup), and the third is burnup measured through measurement equipment (measured burnup). This paper describes a comparative evaluation of design burnup, actual burnup, and measured burnup for specific fuels (40 bundles).