• Proceedings of the Earthquake Engineering Society of Korea Conference
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• 2005.03a
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• pp.55-61
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• 2005

The seismic response tests of a spent fuel dry storage cask model of 1/8 scale are performed for an typical 1940 Elcentro earthquake. This paper focuses on the seismic response test data generation to check the overturing possibility of a storage cask and the slipping displacement on concrete slab bed. A simplified cask model is used to take into account the variations in seismic load magnitude and cask/bed interface friction. The test results show that the model gives an overturning response for an extreme condition.

• Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT)
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• v.18 no.3
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• pp.395-405
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• 2020

Compartment fire tests were performed using kerosene and Jet A-1 as fire sources to evaluate the relationship between flame temperature and opening size. The tests were performed for a fire caused by the release of kerosene owing to vehicle impact, and for a fire caused by the release of Jet-A-1 owing to airplane collision. The compartment fire tests were performed using a 1/3-scale model of a metal storage cask when the flame temperature was deemed to be the highest. We found the combustion time of Jet-A-1 to be shorter than that of kerosene, and consequently, the flame temperature of Jet-A-1 was measured to be higher than that of kerosene. When the opening was installed on the compartment roof, even though the area of the opening was small, the ventilation factor was large, resulting in a high flame temperature and long combustion. Therefore, the position of the opening is a crucial factor that affects the flame temperature. When the metal storage cask was stored in the compartment, the flame temperature decreased proportionally with the energy that the metal storage cask received from the flame.

• Nuclear Engineering and Technology
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• v.53 no.7
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• pp.2268-2276
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• 2021

Spent nuclear fuel, which can degrade during long-term storage, must be transported intact in normal transport conditions. In this regard, many studies, including those involving Multi-Modal Transportation Test (MMTT) campaigns, have been conducted. In order to transport the spent fuel safely, a tie-down structure for supporting and transporting a cask containing the spent fuel is essential. To ensure its structural integrity, a method for finding an optimum conceptual design for the tie-down structure is presented. An optimized transportation test model of a tie-down structure for the KORAD-21 metal cask is derived based on the proposed optimization approach, and the transportation test model is manufactured by redesigning the optimized model to enable its producibility. The topology optimization approach presented in this paper can be used to obtain optimum conceptual designs of tie-down structures developed in the future.

• Nuclear Engineering and Technology
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• v.55 no.7
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• pp.2604-2612
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• 2023

Spent nuclear fuel and long-lived radioactive waste must be carefully handled before disposing them off to a geological repository. After the pre-storage period in water pools, spent nuclear fuel is stored in casks, which are widely used for interim storage. Interim storage in casks is very important part in the whole cycle of nuclear energy generation. This paper presents the results of the numerical study that was performed to evaluate the thermal behavior of a metal-concrete CONSTOR RBMK-1500 cask loaded with spent nuclear fuel and placed in an open type interim storage facility which is under fire conditions (steady-state, fire, post-fire). The modelling was performed using the ANSYS Fluent code. Also, a local sensitivity analysis of thermal parameters on temperature variation was performed. The analysis demonstrated that the maximum increase in the fuel load temperatures is about 10 ℃ and 8 ℃ for 30 min 800 ℃ and 60 min 600 ℃ fires respectively. Therefore, during the fire and the post-fire periods, the fuel load temperatures did not exceed the 300 ℃ limiting temperature set for an RBMK SNF cladding for long-term storage. This ensures that fire accident does not cause overheating of fuel rods in a cask.

• Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT)
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• v.2 no.1
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• pp.13-22
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• 2004

This study presents the thermal analyses of a spent fuel dry storage cask under normal and off-normal conditions. The environmental temperature is assumed to be 15 $^{\circ}C$ under the normal condition. The off-normal condition has an environmental temperature of 38 $^{\circ}C$. An additional off-normal condition is considered as a partial blockage of the air inlet ducts. Two of the four air inlet ducts are assumed to be completely blocked. The significant thermal design feature of the storage cask is the air flow path used to remove the decay heat from the spent fuel. Natural circulation of the air inside the cask allows the concrete and fuel cladding temperatures to be maintained below the allowable values. The finite volume computational fluid dynamics code FLUENT was used for the thermal analysis. The maximum temperatures of the fuel rod and concrete overpack were lower than the allowable values under the normal and off-normal conditions.

• Nuclear Engineering and Technology
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• v.51 no.8
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• pp.2018-2025
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• 2019

This study systematically compared two hybrid deterministic/Monte Carlo transport codes, ADVANTG/MCNP and MAVRIC, in solving a difficult shielding problem for a real-world spent fuel storage cask. Both hybrid codes were developed based on the consistent adjoint driven importance sampling (CADIS) methodology but with different implementations. The dose rate distributions on the cask surface were of primary interest and their predicted results were compared with each other and with a straightforward MCNP calculation as a baseline case. Forward-Weighted CADIS was applied for optimization toward uniform statistical uncertainties for all tallies on the cask surface. Both ADVANTG/MCNP and MAVRIC achieved substantial improvements in overall computational efficiencies, especially for gamma-ray transport. Compared with the continuous-energy ADVANTG/MCNP calculations, the coarse-group MAVRIC calculations underestimated the neutron dose rates on the cask's side surface by an approximate factor of two and slightly overestimated the dose rates on the cask's top and side surfaces for fuel gamma and hardware gamma sources because of the impact of multigroup approximation. The fine-group MAVRIC calculations improved to a certain extent and the addition of continuous-energy treatment to the Monte Carlo code in the latest MAVRIC sequence greatly reduced these discrepancies. For the two continuous-energy calculations of ADVANTG/MCNP and MAVRIC, a remaining difference of approximately 30% between the neutron dose rates on the cask's side surface resulted from inconsistent use of thermal scattering treatment of hydrogen in concrete.

• Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT)
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• v.4 no.2
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• pp.139-152
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• 2006

A spent fuel storage cask is required to prove the safety of a canister under a hypothetical accidental drop condition. A hypothetical accidental drop condition means that a canister is assumed to be a lee drop on to a pad of the storage cask during loading it into a storage cask. A pad of the storage cask absorbs shock to maintain the structural integrities of a canister under a hypothetical accidental drop condition. In this paper a finite element analysis for various pad structures was carried out to improve the structural integrity of a canister under a hypothetical accidental drop condition. A pad of a storage cask was designed a steel structure with concrete. The 1/4 height of a pad was modified with a structure composed of a steel and a polyurethane foam as a impact limiter. The effect of a shape of a steel structure was studied. The effects of the thickness of a steel structure and the density of a polyurethane foam was also studied.

• Journal of computational fluids engineering
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• v.21 no.4
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• pp.33-39
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• 2016

Generally, thermal analysis of spent fuel storage cask has been conducted using the porous media and effective thermal conductivity model to simplify the structural complexity of spent fuel assemblies. As the fuel assembly is composed of two regions; active fuel region corresponding to UO2 pellets and unactive fuel region corresponding to the top and bottom nozzle, the heat transfer performance can be influenced depending on porous media application at these regions. In this study, numerical analysis on concrete storage cask of spent fuel was performed to investigate heat transfer effects for two cases; one was porous media application only to active fuel region(case 1) and the other one was porous media to whole length of fuel assembly(case 2). Using computational fluid dynamics code, the three dimensional, 1/4 symmetry model was constructed. For two cases, maximum temperatures for each component were evaluated below the allowable limits. For the case 1, maximum temperatures for fuel cladding, neutron absorber and baskets inside the canister were slightly higher than those for the case 2. In particular, even though the helium flows with low velocity due to buoyant forces occurred at the top and bottom of unactive fuel region, treating only active fuel region as the porous media was ineffective in respect of the heat removal performance of concrete storage cask, implying a conservative result.

• Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT)
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• v.4 no.3
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• pp.245-253
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• 2006

A test of the 1/3 scale model was conducted to verify the tip-over analysis of a dry. concrete storage cask under a hypothetical accident condition. The tip-over analysis was executed using the velocity at each point as the initial conditions of the model just before the impact. The initial velocity was determined from the initial angular velocity, which would make the equivalent kinetic energy to the potential energy. To confirm the structural integrity of the canister, the visual testing and the non-detective testings such as Liquid Penetrant testing and Ultrasonic Testing were conducted. The lid of a storage cask was plastically deformed near the impact point. The structural integrity of storage cask was maintained. To verify the tip-over analysis the strains and the accelerations acquired by the tip-over test were compared with those by the analyses. The results of the analysis were larger than the test results about two times.

• Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT)
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• v.14 no.4
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• pp.411-422
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• 2016

Recently, because the wet pool storage facilities of NPPs in Korea has become saturated, there has been much active R&D on an interim dry storage system using a transportation and storage cask. Generally, the shielding evaluation for the design of a spent fuel transportation and storage cask is performed by the design basis fuel, which selects the most conservative fuel among the fuels to be loaded into the cask. However, the loading of actual spent fuel into the transportation metal cask is not limited to the design basis fuel used in the shielding evaluation; the loading feasibility of actual spent fuel is determined by the shielding evaluation that considers the characteristics of the initial enrichment, the maximum burnup and the minimum cooling period. This study describes a shielding analysis method for determining the minimum cooling period of spent fuel that meets the domestic transportation standard of the dual purpose metal cask. In particular, the spent fuel of 3.0~4.5wt% initial enrichment, which has a large amount of release, was evaluated by segmented shielding calculations for efficient improvement of the results. The shielding evaluation revealed that about 81% of generated spent fuel from the domestic nuclear power plants until 2008 could be transported by the dual purpose metal cask. The results of this study will be helpful in establishing a technical basis for developing operating procedures for transportation of the dual purpose metal cask.