• Transactions of the Korean Society of Mechanical Engineers A
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• v.21 no.3
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• pp.457-469
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• 1997

A nuclear spent fuel shipping cask is required by IAEA and domestic regulations to withstand a 9m free drop condition. In this paper, the structural analysis under the 9m side drop condition was performed to understand the dynamic impact behavior and to evaluate the safety of the cask for 7 PWR nuclear spent fuel assemblies. The analysis result was compared with the measured value of the 9m side drop test for the 1/3 scaled-down model and the accuracy of the 3D analysis was confirmed. Analysis in accordance with the diameter of impact limiters for the proto-type cask were performed. Through the analysis, the impact behaviors due to the side drop and the effects dependent on the diameter of impact limiters were grasped. Maximum stress intensities on each part of the cask were respectively calculated by using the stress evaluation program and the structural safety of the cask was finally evaluated in accordance with the regulations.

• Journal of the Computational Structural Engineering Institute of Korea
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• v.18 no.4 s.70
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• pp.373-383
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• 2005

The cask is used to transfer the radioactive material in various fields required to withstand hypothetical accident condition such as 9m drop impact in accordance with the requirement of the domestic requlations and IAEA. So far the impact force has been obtained by the finite element method with complex computational procedure. In this study, the dynamic impact response of the cask body is analyzed using the mode superposition method, and the analysis method is proposed. The results we also validated by comparing with previous experimental results and finite element analysis results. The present method Is simpler than finite element method and can be used to predict the global impact response of cask

• Nuclear Engineering and Technology
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• v.48 no.2
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• pp.505-517
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• 2016

The structural integrity of a dual-purpose metal cask currently under development by the Korea Radioactive Waste Agency (KORAD) was evaluated, through numerical simulations and a model test, under high-speed missile impact reflecting targeted aircraft crash conditions. The impact conditions were carefully chosen through a survey on accident cases and recommendations from literature. In the impact scenario, a missile flying horizontally hits the top side of the cask, which is freestanding on a concrete pad, with a velocity of 150 m/s. A simplified missile simulating a commercial aircraft engine was designed from an impact loade-time function available in literature. In the analyses, the dynamic behavior of the metal cask and the integrity of the containment boundary were assessed. The simulation results were compared with the test results for a 1:3 scale model. Although the dynamic behavior of the cask in the model test did not match exactly with the prediction from the numerical simulation, other structural responses, such as the acceleration and strain history during the impact, showed very good agreement. Moreover, the containment function of the cask survived the missile impact as expected from the numerical simulation. Thus, the procedure and methodology adopted in the structural numerical analyses were successfully validated.

• Nuclear Engineering and Technology
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• v.51 no.1
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• pp.284-292
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• 2019

For the continuous operation of a nuclear reactor, burnt fuel needs to be replaced with fresh fuel, where appropriate (ex-vessel) fuel handling is required. Particularly for the Sodium-cooled Fast Reactor (SFR) refueling, its process has unique challenges due to liquid sodium coolant. The ex-vessel spent fuel transportation should concern several design features such as the radiation shielding, decay-heat removal, and inert space separated from air. This paper proposes a new design optimization methodology of cask shielding to transport the spent fuel assembly in a prototype SFR for the first time. The Particle Swarm Optimization (PSO) algorithm had been applied to design trade-offs between shielding and cask weight. The cask is designed as a double-cylinder structure to block an inert sodium region from the air-cooling space. The PSO process yielded the optimum shielding thickness of 26 cm, considering the weight as well. To confirm the shielding performance, the radiation dose of spent fuel removed at its peak burnup and after 1-year cooling was calculated. Two different fuel positions located during transportation were also investigated to consider a functional disorder in a cask drive system. This study concludes the current cask design in normal operations is satisfactory in accordance with regulatory rules.

• Nuclear Engineering and Technology
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• v.54 no.10
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• pp.3766-3777
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• 2022

It is essential to ensure the safety of spent nuclear fuel (SNF) transport cask in drop situation that is included in transport accident scenarios. The safety of the drop situation is affected by the impact absorption performance of impact limiters. Therefore, when designing an impact limiter, the uncertainty in the material properties that affect the impact absorption performance must be considered. In this study, the material properties of the wood inside the impact limiter were selected as the variables for a parametric study. The sensitivity analysis of the drop response of the SNF transport cask with impact limiter was performed. The minimum wood strength required to prevent a direct collision between the cask and floor was derived from the analysis results. In addition, the plastic strain response was analyzed and strain-based evaluation was performed. Based on this result, the critical values of wood properties that change the impact dynamic characteristics were investigated. Finally, the optimal material properties of wood were obtained to secure the structural safety of the SNF transport cask. The results of this study can contribute to the development of SNF transport cask, thereby ensuring safety in transport accident 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.

• Nuclear Engineering and Technology
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• v.34 no.3
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• pp.187-201
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• 2002

The KN-12 cask is designed to transport 12 PWR spent nuclear fuels and to comply with the requirements of Korea Atomic Energy Act, IAEA Safety Standards Series No.57-1 and US 10 CFR Part 71 for a Type B(U)F package. It provides containment, radiation shielding, structural integrity, criticality control and heat removal for normal transport and hypothetical accident conditions. W.H 14$\times$14, 16$\times$16 and 17$\times$17 fuel assemblies with maximum allowable initial enrichment of 5.0 wt.%, maximum average burn-up of 50,000 MWD/MTU and minimum cooling time of 7 years being used in Korea will be loaded and subsequently transported under dry and wet conditions. A forged cylindrical cask body which constitutes the containment vessel is closed by a cask lid. Polyethylene rods for neutron shielding are arranged in two rows of longitudinal bore holes in the cask body wall. A fuel basket to accommodate up to 12 PWR fuel assemblies provides support of the fuels, control of criticality and a path to dissipate heat. Impact limiters to absorb the impact energy under the hypothetical accident conditions are attacked at the top and at the bottom side of the cask during transport. Handling weight loaded with water is 74.8 tons and transport weight loaded with water with the impact limiters is 84.3 tons. The cask will be licensed in accordance with Korea Atomic Energy Act 3nd fabricated in Korea in accordance with ASME B&PV Code Section 111, Division 3.

• 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.

• Proceedings of the Korean Society for Technology of Plasticity Conference
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• 2002.05a
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• pp.98-102
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• 2002

As the nuclear power plant has been operated continuously and increased gradually, transportation and storage of spent fuel are seriously considered nowadays. The transportation cask which contains radioactive material needs to be inspected about structural safety. About safety verification, description of IAEA Safety Standards states that cask must withstand hypothetical accident conditions. In this paper, 9m free drop impact analysis was performed for transportation cask and impact limiter by using the finite element methods. Furthermore, we obtained the dynamic behavior of wood to as compared with safety test results, and verified the safety of transportation cask.

• Transactions of the Korean Society of Pressure Vessels and Piping
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• v.18 no.2
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• pp.50-53
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• 2022

To investigate the validity of the finite element analysis program to assess structural integrity of a spent nuclear fuel transport cask subjected to extreme impact loads, structural integrity of the cask for the case of an aircraft engine collision is evaluated using three FE analysis programs: Autodyn, Speed and ABAQUS explicit version. As a result of all analyses, it is confirmed that no penetration occurred in the cask wall. Even though the different programs are used, it is identified that there are insignificant differences in the FE analysis variables such as von Mises effective stress and equivalent plastic strain among the programs.