The aim of this study is to ensure the structural integrity of a canister to be used in a dry storage system currently being developed in Korea. Based on burnup and cooling periods, the canister is designed with 24 bundles of spent nuclear fuel stored inside it. It is a cylindrical structure with a height of 4,890 mm, an internal diameter of 1,708 mm, and an inner length of 4,590 mm. The canister lid is fixed with multiple seals and welds to maintain its confinement boundary to prevent the leakage of radioactive waste. The canister is evaluated under different loads that may be generated under normal, off-normal, and accident conditions, and combinations of these loads are compared against the allowable stress thresholds to assess its structural integrity in accordance with NUREG-2215. The evaluation result shows that the stress intensities applied on the canister under normal, off-normal, and accident conditions are below the allowable stress thresholds, thus confirming its structural integrity.

The Wolsong Nuclear Power Plant (NPP) operates an on-site spent fuel dry storage facility using concrete silo and vertical module systems. This facility must be safely maintained until the spent nuclear fuel (SNF) is transferred to an external interim or final disposal facility, aligning with national policies on spent nuclear fuel management. The concrete silo system, operational since 1992, requires an aging management review for its long-term operation and potential license renewal. This involves comparing aging management programs of different dry storage systems against the U.S. NRC’s guidelines for license renewal of spent nuclear fuel dry storage facilities and the U.S. DOE’s program for long-term storage. Based on this comparison, a specific aging management program for the silo system was developed. Furthermore, the facility’s current practices—periodic checks of surface dose rate, contamination, weld integrity, leakage, surface and groundwater, cumulative dose, and concrete structure—were evaluated for their suitability in managing the silo system’s aging. Based on this review, several improvements were proposed.

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.

Since the September 11 terrorist attacks in the United States, concerns about intentional aircraft crashes into nationally critical facilities have soared in countries around the world. The United States government advised nuclear utilities to strengthen the security of nuclear power plants against aircraft crashes and stipulated aircraft crash assessment for new nuclear facilities. Interest in military missile attacks on nuclear facilities has grown after Russia attacked Ukraine’s Zaporizhzhia nuclear power plant, where spent nuclear power dry storage facility is operated. Spent nuclear fuel dry storage facilities in nuclear power plant sites should also strengthen security in preparation for such aircraft crashes. Most, but not all, spent nuclear fuel dry storage facilities in Europe, Japan and Canada are operated within buildings, while the United States and Korea operate dry storage facilities outdoors. Since all of Korea’s dry storage systems are concrete structures vulnerable to crash loads and are exposed to the outside, it is more necessary to prepare for aircraft crash terrorist attacks due to the Korea’s military situation. Residents near nuclear power plants are also demanding assessment and protective measures against such aircraft crashes. However, nuclear power plants, including spent nuclear fuel dry storage facilities, are strong structures and have very high security, so they are unlikely to be selected as targets of terrorism, and spent nuclear fuel dry storage systems are so small that aircraft cannot hit them accurately. Collected opinions on the assessment of aircraft crash accidents at spent nuclear fuel dry storage facilities in nuclear power plant sites were reviewed. In addition, the current laws and regulatory requirements related to strengthening the security of new and existing nuclear power plants against intentional aircraft crashes are summarized. Such strengthening of security can not only ensure the safety of on-site spent nuclear fuel dry storage facilities, but also contribute to the continuous operation of nuclear power plants by increasing resident acceptance.

The Comprehensive Analyzer of Real Estimation for spent fuel POOL (CAREPOOL) has been developed for evaluating the thermal safety of a spent nuclear fuel pool (SFP) during the normal and accident conditions. The management of spent nuclear fuel function provides a management tool for spent nuclear fuel in the SFP. The fuel assemblies both in SFP and reactor side can be shown graphically in the screen. The loading sequence into transfer cask can be checked respectively in the CAREPOOL. A basic heat balance equation was used to estimate the SFP temperature using the heat load calculated in the previous step. The characteristics of typical SFPs and associated cooling systems at reactor sites in the Korea were applied. Accident simulation like station black out leading to loss of SFP cooling or inventory is possible. Emergency cooling water injection pipe installed subsequent to the events at Fukushima 2011 is also modeled in this system. The CAREPOOL provides four main functions- management of spent nuclear fuel, decay heat calculation by ORIGEN-S code, estimation of the time to boil/fuel uncovering by thermal-hydraulics calculations, fuel selection for periodic spent fuel transferring campaign. All of these are integrated into the GUI based CAREPOOL system. The CAREPOOL would be very beneficial to nuclear power plant operator and trainee who have responsibility for the SFP operation.

The dry storage of spent fuel has become an increasingly important issue in the field of nuclear energy. Square-gridded baskets have been widely used for the storage of spent fuel because of their superior heat transfer and structural integrity. In this paper, we review the fabrication process of square-gridded baskets for dry storage of spent fuel. The review includes the design considerations, material selection, manufacturing methods, and quality control measures. We also discuss the challenges and opportunities for further improvement in the fabrication of square-gridded baskets. The fabrication of square-gridded baskets is a critical process for the safe and reliable dry storage of spent fuel. The review of the fabrication process highlights the importance of design considerations, material selection, manufacturing methods, and quality control measures. Continued efforts to improve the fabrication process will help to ensure the safe and secure storage of spent fuel.

The spent fuel is classified based on the arrangement of fuel rods, which is considered the primary characteristic data for selecting nuclear fuel. The reason for prioritizing the classification by fuel rod arrangement is that it has the greatest physical impact on the production, supply, operation, reactor type, rack size within the containment vessel, and specifications for the basket in the future dry storage system. Additionally, as mentioned earlier, various meanings of nuclear fuel types are distinguished according to the arrangement of fuel rod. The burnup and cooling period ranges are also important factors in the characterization analysis for the selection of spent fuel, the burnup range was set for both low and high burnup ranges and the cooling period is necessary to consider the reliability during handling of nuclear fuel thermal distribution within the storage system

On-site storage facility using concrete silo dry storage systems for spent nuclear fuel at Wolsong NPP site came into operation in 1992 and was expanded four times, and a total of 300 silo dry storage systems are currently in operation. The design lifetime of silo dry storage systems has been licensed for 50 years. As the dry storage systems are subject to time constraints for a limited lifetime, countries operating the dry storage systems are working to ensure the long-term integrity of dry storage systems and IAEA also recommends that the dry storage systems be assessed for long-term storage. To demonstrate the long-term integrity due to material degradation during the licensed design lifetime, the structural integrity of silo dry storage systems was evaluated by considering the material degradation characteristics of concrete. The concrete compressive strength results measured so far by the rebound hammer method, which is an internationally standardized nondestructive test method for converting hardness into compressive strength using the correlation between rebound number and strength at the time of a Schmidt hammer strike, were analyzed in accordance with Wolsong NPP’s procedure to quantify the degradation characteristics, and the prediction of concrete strengths for 20 years and 50 years after construction of the silo dry storage systems was determined, respectively. Based on these residual compressive strengths, structural analyses of the silo dry storage systems were carried out under normal, off-normal and accident conditions of the related regulations, and the structural integrity of silo dry storage systems was reevaluated. It was confirmed the silo dry storage systems are able to maintain structural integrity up to the design lifetime of 50 years even if the concrete is deteriorated.

Spent fuel from the Wolsong CANDU reactor has been stored in above-ground dry storage canisters. Wolsong concrete dry storage canisters (silos) are around 6 m high, 3 m in outside diameter, and have shielding comprised of around 1 m of concrete and 10 mm of steel liner. The storage configuration is such that a number of fuel bundles are placed inside a cylindrical steel container known as a Fuel Basket. The canisters hold up to 9 baskets each that are 304 L stainless steel, around 42” in diameter, 22” in height, and hold 60 fuel bundles each. The operating license for the dry storage canisters needs to be extended. It is desired to perform in-situ inspections of the fuel baskets to very their condition is suitable for retrieval (if necessary) and that the temperature within the fuel baskets is as predicted in the canister’s design basis. KHNP-CNL (Canadian Nuclear Lab.) has set-up the design requirements to perform the in-situ inspections in the dry storage canisters. This Design Requirements applies to the design of the dry storage canister inspection system.

PWR spent nuclear fuel generally showed an oxide film thickness of 100 um or more with a combustion rate of 45 MWD/MTU or higher, while CANDU spent nuclear fuel with an average combustion rate of about 7.8 MWD/MTU had few issues related to hydride corrosion. Even based on the actual power plant data, it is known that the thickness of the oxide film is 10 μm or less on the surface of the coating tube, and brittleness caused by hydride is shown from the thickness of the oxide film of about 80 μm, so it is not worth considering. However, since corrosion may be accelerated by lithium ions, lithium ions may be said to be a very important factor in controlling the hydro-chemical environment of heavy water. Lithium has a negative effect on the corrosion of zirconium alloys. However, since local below 5 ppb to prevent corrosion. maintained at a concentration between 0.35 and 0.55 ppm. Hydrogen is known to have a positive effect by suppressing radioactive decomposition of the coolant and suppressing cracks in nickelbased alloys. However, too much hydrogen can produce hydride in a pressure tube composed of Zr-2.5Nb, so DH (Disolved Hydrogen) maintains the range of 0.27–0.90 ppm. pH and conductivity are completely determined by lithium ions, and DH can be completely removed below 5 ppb to prevent corrosion. Therefore, for cladding corrosion simulation of the CANDU spent nuclear fuel, a hydrochemical of the equipment, not 310°C, and 14 uS·Cm−1 is targeted as conditions for corrosion acceleration. In addition, for acceleration, the temperature was set to 345°C (margin 10°C), which is the maximum accommodation range of the equipment, not 310°C.