• Nuclear Engineering and Technology
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• v.52 no.5
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• pp.1079-1092
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• 2020

A parametric study of several parameters relevant to design safety on the spent nuclear fuel (SNF) rod response under a drop accident is presented. In the view of the complexity of interactions between the independent safety-related parameters, a factorial design of experiment is employed as an efficient method to investigate the main effects and the interactions between them. A detailed single full-length fuel rod is used with consideration of post-irradiated fuel conditions under horizontal and vertical free-drops onto an unyielding surface using finite-element analysis. Critical drop heights and critical g-loads that yield the threshold plastic strain in the cladding are numerically estimated to evaluate the fuel rod structural resistance to impact load. The combinatory effects of four uncertain parameters (pellet-cladding interfacial bonding, material properties, spacer grid stiffness, rod internal pressure) and the interactions between them on the fuel rod response are investigated. The principal finding of this research showed that the effects of above-mentioned parameters on the load-carrying capacity of fuel rod are significantly different. This study could help to prioritize the importance of data in managing and studying the structural integrity of the SNF.

• Nuclear Engineering and Technology
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• v.53 no.8
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• pp.2682-2695
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• 2021

This paper aims to evaluate the structural dynamic responses and damage/failure of the nuclear fuel reprocessing plant under the free drop impact of spent fuel cask (SFC) and fuel assembly (FA) during the on-site transportation. At the present Part I of this paper, the large-scale SFC model free drop test and the corresponding numerical simulations are performed. Firstly, a composite target which is composed of the protective structure, i.e., a thin RC plate (representing the inverted U-shaped slab in the loading shaft) and/or an autoclaved aerated concrete (AAC) blocks sacrificial layer, as well as a thick RC plate (representing the bottom slab in the loading shaft) is designed and fabricated. Then, based on the large dropping tower, the free drop test of large-scale SFC model with the mass of 3 t is carried out from the height of 7 m-11 m. It indicates that the bottom slab in the loading shaft could not resist the free drop impact of SFC. The composite protective structure can effectively reduce the damage and vibrations of the bottom slab, and the inverted U-shaped slab could relieve the damage of the AAC blocks layer dramatically. Furthermore, based on the finite element (FE) program LS-DYNA, the corresponding refined numerical simulations are performed. By comparing the experimental and numerical damage and vibration accelerations of the composite structures, the present adopted numerical algorithms, constitutive models and parameters are validated, which will be applied in the further assessment of drop impact effects of full-scale SFC and FA on prototype nuclear fuel reprocessing plant in the next Part II of this paper.

• Proceedings of the Korean Radioactive Waste Society Conference
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• 2004.02a
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• pp.191-204
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• 2004

The research and development of effective management technologies of the spent fuels discharged from power reactors are an important and essential task of KAERI. In resent several years KAERI has focused on a project named "development and demonstration of the Advanced spent fuel Conditioning Process (ACP) in a laboratory scale." The Facility for ACP demonstration consists of two Hot Cells and auxiliary facilities. It is now in the final design stage and will be constructed in 2004. After construction of the facility the ACP equipments will be installed in Hot Cells. The ACP will be demonstrated by some simulated spent fuels first and then by spent fuels.

• Journal of Radiation Protection and Research
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• v.28 no.4
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• pp.361-368
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• 2003

The economic and environmental friendliness analysis of the nuclear fuel cycle options that can be expected in Korea were performed. Options considered are direct disposal, reprocessing and DUPIC (Direct Use of Spent PWR Fuel In CANDU Reactors). By considering the result of calculation of the annual uranium requirement and nuclear spent fuel generation by analysis of nuclear fuel material flows in the nuclear fuel cycle options, we decided the time of back-end nuclear fuel cycle processes and the volume. Then we can analyze the economic and environmental friendliness by applying the unit cost and unit value of each process, respectively.

• Nuclear Engineering and Technology
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• v.54 no.9
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• pp.3353-3360
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• 2022

Using data from the design earthquake of the Korean standard nuclear power plant, seismic analyses of a fuel assembly are conducted in this study. The modal characteristics are used to develop an input deck for the seismic analysis. With a time history analysis, the responses of the fuel assembly in the event of an earthquake are obtained. In particular, the displacement, velocity, and acceleration responses at the center location of the fuel assembly are obtained in the time domain, with these outcomes then used for a detailed structural analysis of the fuel rods in the ensuing analyses. The response spectra are also generated to determine the response characteristics in the frequency domain. The structural integrity of the fuel assembly can be ensured through this type of time history analysis considering the input excitations of various earthquakes considered in the design.

• Advances in environmental research
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• v.6 no.4
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• pp.301-315
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• 2017

An Amberlite XAD-4 resin impregnated with di(2-ethylhexyl)phosphoric acid was prepared and its adsorption-desorption behaviors with Sr(II) ions under various conditions was examined. The resin was characterized by fourier transform infrared and thermal analysis techniques. The effects contact time, temperature, pH, interfering ions and eluants were studied. Results showed that adsorption of Sr (II) well fitted with pseudo-second-order kinetic model. The equilibrium adsorption data of Sr (II) on the impregnated resin were analyzed by Jossens, Weber-van Vliet, Redlich-Peterson and Fritz-Schlunder models to find out desirable equilibrium condition. Among them, the Fritz-Schlunder model best fitted to the experimental data. The maximum sorption capacity of impregnated resin amounted to 0.45 mg/ g at pH 8.0 and $20^{\circ}C$.

• Nuclear Engineering and Technology
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• v.38 no.5
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• pp.417-422
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• 2006

The present paper describes the 45 MW High Flux Reactor (HFR) which is located in Petten, The Netherlands. This paper focuses on selected technical aspects of this reactor and on nuclear fuel irradiation experiments. These fuel experiments are mainly experiments on Light Water Reactor (LWR) and Very/High Temperature Reactor (V/HTR) fuels, but also on Fast Reactor (FR) fuels, transmutation fuels and Material Test Reactor (MTR) fuels.

• Nuclear Engineering and Technology
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• v.42 no.1
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• pp.125-130
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• 2010

Sipping measurements were implemented at the Open Pool Australian Light water reactor (OPAL) to test for failure in reactor fuel elements. Fission product released by the fuel element into the pool water was measured using both High Purity Germanium (HPGe) detection via samples and a NaI(Tl) detection in-situ with the sipping device. Results from two fuel elements are presented.

• Nuclear Engineering and Technology
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• v.50 no.5
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• pp.717-723
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• 2018

Thermal analysis of accident conditions is an important problem during safety assessment of the dry spent nuclear fuel storage facilities. Thermal aspects of accident conditions with channel blockage of ventilated storage containers are considered in this article. Analysis of flow structure inside ventilated containers is carried out by numerical simulation. The main mechanisms of heat and mass transfer, which take part in spent nuclear fuel cooling, were detected. Classification of accidents on the basis of their influence on the maximum temperatures inside storage casks is proposed.

• Nuclear Engineering and Technology
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• v.37 no.4
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• pp.299-308
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• 2005

The behavior of fission gas in all major types of nuclear fuel has been reviewed with an emphasis on more recently discovered aspects. It is proposed that the behavior of fission gas can be classified in a number of characteristic types that occur at a high or low operating temperature, and/or at high or low fissile burnup. The crystal structure and microstructure of the various fuels are the determinant factors in the proposed classification scheme. Three types of behavior, characterized by anisotropic $\alpha$-U, high temperature metallic $\gamma$-U, and cubic ceramics, are well-known and have been extensively studied in the literature. Less widely known are two equally typical low temperature kinds: one associated with fission induced grain refinement and the other with fission induced amorphization. Grain refinement is seen in crystalline fuel irradiated to high burnup at low temperatures, whereas breakaway swelling is observed in amorphous fuel containing sufficient excess free-volume. Amorphous fuel, however, shows stable swelling if insufficient excess free-volume is available during irradiation.