Due to the saturation of spent fuel pool of nuclear power plant in Korea, temporary storage for spent fuel will be installed, and spent fuel will be stored and managed in dry cask for a considerable period of time. Since spent nuclear fuel must withstand continuous decay heat, radiation and high internal pressure of the fuel rod in the cask, behavior of spent nuclear fuel is needed to be reviewed. Spent nuclear fuel used in Pressurized Water Reactor (PWR) in Korea is stored in a wet storage currently, but it is going to store a temporary dry-storage facility on Kori site. Therefore, it is very important and meaningful to evaluate the behavior of nuclear fuel with realistic modeling. Also, domestic PWR nuclear fuel has various burn-up. In the past, the burn-up of nuclear fuel in light water reactors was low, but in order to increase power generation efficiency, the concentration of uranium was increased and the number of new fuel was increased. Therefore, a large amount of nuclear fuel with burn-up of 45,000 MWD/MTU or higher, generally called high burn-up, is also stored in the spent fuel pool (SFP). Therefore, it is necessary to evaluate by dividing three different burn-up such as, low, medium, and high burn-up. Thus, this study will review the behavior of nuclear fuel at different burn-up during the temporary storage period with FALCON (EPRI), computational code and analyze the factors affecting the integrity of nuclear fuel, including when the temporary storage is extended its additional lifetime. And this evaluation will contribute developing the spent fuel management plan in Korea.

In KNF, fuel performance analysis modules were developed to predict the overall behavior of a fuel rod under normal operating conditions. Their main focus is to provide information on initial conditions prior to dry storage. Potential degradation mechanisms that may affect sheath integrity of spent CANDU fuel during dry storage are: creep rupture under internal gas pressure, sheath oxidation in air environment, stress corrosion cracking, delayed hydride cracking, and sheath splitting due to UO2 oxidation for a defective fuel. To upgrade the developed modules that address all the damage mechanisms, the first step was a review of the available technical information on phenomena relevant to fuel integrity. The second step was an examination of the technical bases of all modules, identify and extend the ranges of all modules to required operating ranges. The 300°C spent CANDU fuel sheath temperature metric for dry storage ensures spent CANDU fuel element integrity from the failure mechanisms of creep rupture, oxidation and stress corrosion cracking at a failure probability of 2×10-5 for a dry storage time of 100 years. The 300°C sheath temperature metric for dry storage has relatively a lower failure rate than the target criteria for dry storage of spent LWR fuel. Although different modes of failure were treated separately for simplicity, ignoring possible synergistic effects, these results are conservative because of the conservative assumptions that have been made for evaluating spent fuel element conditions, and because of the inherent conservatism of the applied models. Additional conservatism of the model comes from the fact that isothermal conditions do not prevail in actual storage conditions. Further R&D being considered includes acquisition of new functional models to implement overall fuel behavior evaluation and cover spent CANDU fuel in dry storage, and upgrades of the analysis module to achieve sufficient accuracy in key output parameters. The developed modules provide a platform for research and industrial applications, including the design of fuel behavior experiments and prediction of safe operating margins for spent CANDU fuel.

A lot of CANDU Spent Fuels (CSFs) have been stored in spent nuclear fuel pools and dry storage facilities. In accordance with the enhanced nuclear regulations, the initial characteristics of CSF should be inspected to ensure the integrity of CSF and the reliable operation of storage system before loading it into a cask for long-term dry storage. For the inspections, an initial characteristics measurement equipment was designed, which is used for Pool-Side Examination (PSE) in the spent fuel pool of the pressurized heavy water reactor nuclear power plant. Measurements using the equipment consist of non-contact inspections and contact inspections. The non-contact inspections do not affect CSF integrity, whereas the integrity of CSF can be reduced during the contact inspections under abnormal operating conditions because the probe of equipment may apply specific loads to the CSF. Therefore, the structural integrity evaluations of equipment and CSF are performed using Finite Element (FE) analyses for four combinations based on two abnormal conditions and two probe positions. The used abnormal conditions are the pressing load condition and the scratching load condition, and two probe positions are the center and bottom of the fuel rod in the longitudinal direction, respectively. In this evaluation, the bottoms of the fuel rod or CSF are defined as the regions facing the bottom surface of equipment. The analysis of the pressing load condition is performed by pressing the probe of the equipment in radial direction of the CSF fuel rod. That of the scratching load condition is carried out by applying a specific radial load to the CSF fuel rod using the probe and then applying the load to the surface of the fuel rod while moving axially along the surface. All combinations are analyzed considering geometric, boundary and material non-linearity under the dynamic load, which is dependent on the equipment operating velocity. The stresses of CSF and equipment components were obtained from these analyses. The maximum stress of each component was generated at the combination on the scratching load condition for the bottom position among the four combinations. The obtained maximum stresses are lower than the yield stress for each component material. Also, the CSF is not overturned due to the support plate of the equipment in all analyses. Therefore, the structural integrity and safety of the equipment and the CSF are maintained under abnormal operating conditions during the inspection using the initial characteristic measurement equipment.

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.

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.

Maintaining fuel sheath integrity during dry storage is important. Intact sheath acts as the primary containment barrier for both fuel pellets and fission products over the dry storage periods and during subsequent fuel handling operations. In KNF, in-house fuel performance code was developed to predict the overall behavior of a fuel rod under normal operating conditions. It includes the analysis modules to predict temperature, pellet cracking and deformation, sheath stress and strain at the mid-plane of the pellet and pellet-pellet interfaces, fission gas release and internal gas pressure. The main focus of the code is to provide information on initial conditions prior to dry storage, such as fission gas inventory and its distribution within the fuel pellet, initial volumes of storage spaces and their locations, radial profile of heat generation within the pellet, etc. To upgrade the developed code that address all the damage mechanisms, the first step was a review of the available technical information on phenomena relevant to fuel integrity. Potential degradation mechanisms that may affect sheath integrity of CANDU spent fuel during dry storage are: creep rupture under internal gas pressure, sheath oxidation in air environment, stress corrosion cracking (SCC), delayed hydride cracking (DHC), and sheath splitting due to UO2 oxidation for a defective fuel. The failure by creep rupture, SCC or DHC is in the form of small cracks or punctures. The failure by sheath oxidation or sheath splitting due to UO2 oxidation results in a gross sheath rupture. The second step was to examine the technical bases of all modules of the in-house code, identify and extend the ranges of all modules to required operating ranges. This step assessed the degradation mechanisms for the fuel integrity. The objective of this assessment is to predict the probability of sheath through-wall failure by a degradation mechanisms as a function of the sheath temperature during dry storage. Further improvements being considered include upgrades of the analysis module to achieve sufficient accuracy in key output parameters. The emphasis in the near future will be on validation of the inhouse code according to a rigorous and formal methodology. The developed models provide a platform for research and industrial applications, including the design of fuel behavior experiments and prediction of safe operating margins for CANDU spent fuel.

Prior to the investigations on fuel degradation it is necessary to describe the reference characteristics of the spent fuel. It establishes the initial condition of the reference fuel bundle at the start of dry storage. In a few technology areas, CANDU fuels have not yet developed comprehensive analysis tools anywhere near the levels in the LWR industry. This requires significantly improved computer codes for CANDU fuel design. In KNF, in-house fuel performance code was developed to predict the overall behavior of a fuel rod under normal operating conditions. It includes the analysis modules to predict temperature, pellet cracking and deformation, clad stress and strain at the mid-plane of the pellet and pellet-pellet interfaces, fission gas release and internal gas pressure. The main focus of the code is to provide information on initial conditions prior to dry storage, such as fission gas inventory and its distribution within the fuel pellet, initial volumes of storage spaces and their locations, radial profile of heat generation within the pellet, etc. Potential degradation mechanisms that may affect sheath integrity of CANDU spent fuel during dry storage are: creep rupture under internal gas pressure, sheath oxidation in air environment, stress corrosion cracking, delayed hydride cracking, and sheath splitting due to UO2 oxidation for a defective fuel. To upgrade the developed code that address all the damage mechanisms, the first step was a review of the available technical information on phenomena relevant to fuel integrity. The second step was an examination of the technical bases of all modules of the in-house code, identify and extend the ranges of all modules to required operating ranges. Further improvements being considered include upgrades of the analysis module to achieve sufficient accuracy in key output parameters. The emphasis in the near future will be on validation of the in-house code according to a rigorous and formal methodology. The developed models provide a platform for research and industrial applications, including the design of fuel behavior experiments and prediction of safe operating margins for CANDU spent fuel.

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.

최근 국내에서 육상 및 해상을 통한 소외 정상운반 시 진동 및 충격하중에 대한 사용후핵연료의 건전성 평가 기술 개발이 수행되고 있다. 이와 관련된 국내 연구사례는 전무하여 기존에 진행된 또는 현재 수행중인 해외연구사례를 조사하여 국내 연구에 참고하고자 한다. 2000년 이전 과거 미국의 사용후핵연료의 정상운반 시 진동 및 충격하중 측정 관련 연구현황을 조사 하였고 2009년부터 미국국립연구소 주관으로 실시한 단축가진시험, 콘크리트블럭 트럭운반시험, 다축가진시험에 대해서 조사하였으며 2017년 미국 SNL, 스페인의 ENSA, 한국이 공동으로 수행한 복합운반시험을 상세히 조사하였다. 시험 준비과정, 절차, 가속도 및 변형률 측정결과, 유한요소 및 다물체동역학 해석과정 등이 조사되었다. 각 시험 별로 측정된 변형률 자료를 바탕으로 사용후핵연료 피로곡선과 비교한 결과 손상을 일으키기에는 매우 미미한 정도의 변형률이 발생한다는 초기 결론을 얻었음을 확인하였다. 하지만 현재 결론은 일부 결과만을 검토한 예비 결론으로 상세한 검토가 현재 미국에서 진행 중이다. 미국에서 지금까지 수행한 사용후핵연료의 정상운반조건에서의 진동 및 충격하중 측정과 관련하여 조사된 내용은, 국내 운반환경에서 사용후핵연료의 정상운반시험을 수행할 때 참고할만한 유용한 자료라 판단된다.