• Proceedings of the Korean Radioactive Waste Society Conference
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• 2017.10a
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• pp.123-124
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• 2017

• Nuclear Engineering and Technology
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• v.53 no.11
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• pp.3860-3860
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• 2021

• Nuclear Engineering and Technology
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• v.50 no.2
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• pp.313-317
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• 2018

Licenses for the storage of spent nuclear fuel (SNF) and vitrified highly active waste in casks under dry conditions are limited to 40 years and have to be renewed for prolonged storage periods. If such a license renewal has to be expected since as in accordance with the new site selection procedure a final repository for spent fuel in Germany will not be available before the year 2050. For transport and possible unloading and loading in new casks for final storage, the integrity and the maintenance of the geometry of the cask's inventory is essential because the SNF rod cladding and the cladding of the vitrified highly active waste are stipulated as a barrier in the storage concept. For SNF, the cladding integrity is ensured currently by limiting the hoop stress and hoop strain as well as the maximum temperature to certain values for a 40-year storage period. For a prolonged storage period, other cladding degradation mechanisms such as inner and outer oxide layer formation, hydrogen pick up, irradiation damages in cladding material crystal structure, helium production from alpha decay, and long-term fission gas release may become leading effects driving degradation mechanisms that have to be discussed.

• Nuclear Engineering and Technology
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• v.53 no.11
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• pp.3704-3710
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• 2021

The characteristics of the external radiation on the surface of the casks for spent nuclear fuel (SNF) storage by dry method are investigated for the case when the spatial distribution of SNF in the basket changes due to the destruction of the fuel rod claddings. The surface areas are determined, where the changes in fluxes of neutrons, produced by ²⁴⁴Cm actinide, and γ-quanta, produced by long-lived isotopes, are maximum in the result of the decrease in the height of the SNF area. Concrete (VSC-24) and metal (SC-21) casks are considered as examples. The procedure of periodic measurement of the dose rate of neutrons or γ-quanta at the specified points of the cask surface is proposed for identifying the fuel rod cladding destruction. Under normal operation, the decrease in the dose rate produced by neutrons as the function of SNF storage duration is determined by the half-life of ²⁴⁴Cm, and for γ-quanta - by the half-lives of long-lived SNF isotopes. Consequently, a stepwise change in the dose rate of neutrons or γ-quanta, detected by the measurements, as compared to the previous one, would indicate the destruction of the fuel rod claddings.

• 제어로봇시스템학회:학술대회논문집
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• 1996.10b
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• pp.1060-1063
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• 1996

Remote Cask Grappling and Lid Unbolting Device (RCGLUD) is developed as a dedicated device capable of performing complete procedure of handling nuclear spent fuel transport cask. Since RCGLUD is suspended to an overhead crane, its body should undergo prolonged vibration upon actuation in rotational direction and it becomes difficult to achieve precise grappling of the cask. Therefore, this paper presents an adaptation of input shaping technique to effectively suppress the rotational vibration of RCGLUD and achieve precise positioning in rotational direction. This technique has a practical merit in that it requires only the information on the system natural frequency and the damping ratio. Its performance is verified by both simulation and experimental studies, and revealed that the method is also insensitive to modeling error.

• Nuclear Engineering and Technology
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• v.45 no.7
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• pp.969-978
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• 2013

A full-sized model for the horizontally oriented metal cask containing 21 spent fuel assemblies has been considered to evaluate the internal natural convection behavior within a dry shield canister (DSC) filled with helium as a working fluid. A variety of two-dimensional CFD numerical investigations using a turbulent model have been performed to evaluate the heat transfer characteristics and the velocity distribution of natural convection inside the canister. The present numerical solutions for a range of Rayleigh number values ($3{\times}10^6{\sim}3{\times}10^7$) and a working fluid of air are further validated by comparing with the experimental data from previous work, and they agreed well with the experimental results. The predicted temperature field has indicated that the peak temperature is located in the second basket from the top along the vertical center line by effects of the natural convection. As the Rayleigh number increases, the convective heat transfer is dominant and the heat transfer due to the local circulation becomes stronger. The heat transfer characteristics show that the Nusselt numbers corresponding to $1.5{\times}10^6$ < Ra < $1.0{\times}10^7$ are proportional to 0.5 power of the Rayleigh number, while the Nusselt numbers for $1.0{\times}10^7$ < Ra < $8.0{\times}10^7$ are proportional to 0.27 power of the Rayleigh number. These results agreed well with the trends of the experimental data for Ra > $1.0{\times}10^7$.

• Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT)
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• v.9 no.3
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• pp.161-167
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• 2011

In case that the maximum temperature of any surface readily accessible during transport of a spent nuclear fuel (SNF) transport cask exceeds $85^{\circ}C$ in the absence of insolation under the ambient temperature of $38^{\circ}C$, personnel barriers or transport hood shall be used to prevent people from casual contact with the transport cask surface. Usually the air temperature within the hood and the hood surface temperature are calculated and further utilized as boundary conditions(free stream temperature and external radiation temperature) for thermal evaluation under normal conditions of transport. In this study, these temperatures are derived using the analytical method based on the heat transfer mechanism around the transport cask under transport hood assuming the thermal equilibrium. By comparing the analytical solutions with the results from the detailed calculations with CFD-computer-code FLUENT 12.1 it is verified that the analytical method is still efficient tool to estimate the temperatures and these temperatures can be further used as boundary conditions for thermal evaluation under normal conditions of transport.

• Nuclear Engineering and Technology
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• v.54 no.11
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• pp.4146-4158
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• 2022

Ultra-high performance concrete (UHPC) has been widely utilized in military and civil protective structures to resist intensive loadings attributed to its excellent properties, e.g., high tensile/compressive strength, high dynamic toughness and impact resistance. At present, aiming to improve the defects of the traditional vertical concrete cask (VCC), i.e., the external storage facility of spent fuel, with normal strength concrete (NSC) shield, e.g., heavy weight and difficult to fabricate/transform, the feasibility of UHPC applied in the shield of VCC is numerically examined considering its high radiation and corrosion resistance. Firstly, the finite element (FE) analyses approach and material model parameters of NSC and UHPC are verified based on the 1/3 scaled VCC tip-over test and drop hammer test on UHPC members, respectively. Then, the refined FE model of prototypical VCC is established and utilized to examine its dynamic behaviors and damage distribution in accidental tip-over and end-drop events, in which the various influential factors, e.g., UHPC shield thickness, concrete ground thickness, and sealing methods of steel container are considered. In conclusion, by quantitatively evaluating the safety of VCC in terms of the shield damage and vibrations, it is found that adopting the 300 mm-thick UHPC shield instead of the conventional 650 mm-thick NSC shield can reduce about 1/3 of the total weight of VCC, i.e., about 50 t, and 37% floor space, as well as guarantee the structural integrity of VCC during the accidental drop simultaneously. Besides, based on the parametric analyses, the thickness of concrete ground in the VCC storage site is recommended as less than 500 mm, and the welded connection is recommended for the sealing method of steel containers.

• 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 Radiation Protection and Research
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• v.22 no.2
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• pp.77-83
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• 1997

Epoxy resin-type neutron shielding materials, KNS(Kaeri Neutron Shield)-101, KNS-102, and KNS-103 have been fabricated to be used in spent fuel shipping cask. The base material is epoxy resin, and polypropylene, aluminium hydroxide, and boron carbide are added. These shielding materials offer good fluidity at processing, which makes it possible to apply this resin shield to complicated geometric shapes such as shipping cask. The shielding property of these shielding materials for shipping cask for loading 28 PWR spent fuel assemblies has been evaluated. ANISN code is used to evaluate the shielding property of the shipping cask with the thickness of the three neutron shielding materials greater than 10 cm. As a result of analysis, the maximum calculated dose rate at the radial surface of the cask is determined to be $300{\mu}Sv/h$ and the maximum calculated dose rate at 100 cm from the cask is $97{\mu}Sv/h$. These dose rates remain within allowable values specified in related regulations.