• 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.

• Journal of Radiation Protection and Research
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• v.28 no.4
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• pp.281-290
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• 2003

The KN-12 spent nuclear fuel transport cask, which is a Type B(U) package designed to comply with the requirements of Korea Atomic Energy Act[1], IAEA Safety Standards Series No.TS-R-1[2] and US 10 CFR Part 71[3], is designed for carrying up to 12 PWR spent fuel assemblies in a basket structure. The cask has been licensed in accordance with Korea Atomic Energy Act and was fabricated in Korea in accordance with the requirements of ASME B&PV Sec.III, Div.3[4]. The cask must maintain thermal integrity in accordance with the related regulations and be evaluated to verify that the thermal performance of the cask complies with the regulatory requirements. The temperatures of the cask and components were determined by using finite elements methods with a numerical tool, safety tests using an 1/8 height slice model of the real cask were conducted to demonstrate verification of the numerical tool and methods, and heat transfer tests for normal transport conditions were performed as a fabrication acceptance test to demonstrate the heat transfer capability of the cask.

• Journal of Radiation Protection and Research
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• v.26 no.3
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• pp.207-214
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• 2001

The KN-12 spent nuclear fuel (SNF) transport cask is designed for transportation of up to 12 assemblies and is in standby status for being licensed in accordance with Korea Atomic Energy Act. To evaluate radiation shielding and criticality safety of the KN-12 cask, each case of study was carried out using MCNP4B Code. MCNP code is verified by performing benchmark calculation for the KSC-4 SNF cask designed in 1989. As a result of radiation safety evaluation for the KN-12 cask, calculated dose rates always satisfied the standards at the cask surface, at 2m from the surface in normal transport condition, and at 1 m from the surface in hypothetical accident condition. Maximum dose rate was always arisen on the side of the cask. For normal transport condition, photons primarily contribute to dose rate between two kinds of released sources, neutrons and photons, from spent nuclear fuel but for hypothetical accident condition, contrary case was resulted. The level of calculated dose rate was 27.8% of the limit at the cask surface, 89.3% at 2 m from the cask surface, and 25.1% at 1 m from the cask surface. For criticality analysis, keff resulting from the criticality analysis considering the condition of optimum partial flooding with fresh water is 0.89708(0.00065. The results confirm the standards recommended by all regulations on radiation safety.

• Journal of Radiation Protection and Research
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• v.28 no.4
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• pp.291-298
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• 2003

The KN-12 transport cask has been designed to transport 12 PWR spent nuclear fuel assemblies and to comply with the regulatory requirements for a Type B(U) package. The containment boundary of the cask is defined by a cask body, a cask lid, lid bolts with nuts, O-ring seals and a bolted closure lid. The containment vessel for the cask consists of a forged thick-walled carbon steel cylindrical body with an integrally-welded carbon steel bottom and is closed by a lid made of stainless steel, which is fastened to the cask body by lid bolts with nuts and sealed by double elastomer O-rings. In the cask lid an opening is closed by a plug with an O-ring seal and covered by the bolted closure lid sealed with an O-ring. The cask must maintain a radioactivity release rate of not more than the regulatory limit for normal transport conditions and for hypothetical accident conditions, as required by the related regulations. The containment requirements of the cask are satisfied by maintaining a maximum air reference leak rate of $2.7{\times}10^{-4}ref.cm^3s^{-1}$ or a helium leak rate of $3.3{\times}10^{-4}cm^3s^{-1}$ for normal transport conditions and for hypothetical accident conditions.

• 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.

• Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT)
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• v.20 no.3
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• pp.357-364
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• 2022

The safety of a KTC-360 transport cask, a large-capacity pressurized heavy-water reactor transport cask that transports CANDU spent nuclear fuel discharged from the reactor after burning in a pressurized heavy-water reactor, must be demonstrated under the normal transport and accident conditions specified under transport cask regulations. To confirm the thermal integrity of this cask under normal transport and accident conditions, high-temperature and fire tests were performed using a one-third slice model of an actual KTC-360 cask. The results revealed that the surface temperature of the cask was 62℃, indicating that such casks must be transported separately. The highest temperature of the CANDU spent nuclear fuel was predicted to be lower than the melting temperature of Zircaloy-4, which was the sheath material used. Therefore, if normal operating conditions are applied, the thermal integrity of a KTC-360 cask can be maintained under normal transport conditions. The fire test revealed that the maximum temperatures of the structural materials, stainless steel, and carbon steel were 446℃ lower than the permitted maximum temperatures, proving the thermal integrity of the cask under fire accident conditions.

• Transactions of the Korean Society of Pressure Vessels and Piping
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• v.19 no.2
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• pp.69-74
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• 2023

The paper presents the results investigating the effect of aircraft engine impact location on the intended function evaluation results of spent nuclear fuel transport cask. As a result of the investigation, it is found that the structural integrity is maintained as the maximum accumulated equivalent plastic strain is below the acceptable criterion regardless of the collision location. It is identified that when the aircraft engine collided with the upper part of the transport cask without considering impact limiter the containment performance is weakened compared to when the aircraft engine collided with the central part.

• Nuclear Engineering and Technology
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• v.40 no.4
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• pp.319-326
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• 2008

To transport process wastes efficiently from the Advanced Spent Fuel Conditioning Pyro-process Facility (ACPF) at the Korea Atomic Energy Research Institute (KAERI), a new hot cell cask has been designed based on an existing hot cell padirac transport cask, with not only a neutron absorber for improved shielding capability, but also a docking facility for an easy docking system. In the new hot cell cask, two kinds of materials have been considered as shielding materials, polyethylene and resin. To verify the transport compatibility of the waste and spent fuel for the ACPF, neutron and photon shielding calculations were performed using the MCNPX code. The source term was evaluated by the ORIGEN-ARP code system based on spent PWR fuel. From the calculation, it was found that the maximum surface dose rates of the hot cell cask with the two candidates were estimated within the limit (2 mSv/hr).

• 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.9 no.3
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• pp.191-198
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• 2011

In general, conventional criticality analyses for spent fuel transport/dry storage systems have been performed based on assumption of fresh fuel concerning the potential uncertainties from number density calculation of Transuranic and Fission Products in spent fuel. However, because of economic loss due to the excessive criticality margin, recently the design of transport/dry storage systems with Burnup Credit(BUC) application has been actively developed. The uncertainties in criticality analyses on transport/storage systems with BUC technique show strong dependance upon initial enrichment and burnup rate, whereas those in the conventional criticality evaluation based on fresh fuel assumption do not show such a dependance. In this study, regulatory-required uncertainties of the criticality analyses for BK 26 Cask, which is conceptually designed spent fuel transport cask with BUC corresponding to the limiting circumstances on nuclear power plants in Korea, are evaluated as a function of initial enrichment and burnup rate. Results of this study will be used as basic data for spent fuel loading curve of BK 26 Cask.