• Transactions of the Korean Society of Mechanical Engineers A
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• v.37 no.7
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• pp.841-849
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• 2013

In this study, a 2-step stamping model with an additional 1st stamping tool is proposed to reduce stamping flaws in the curved parts of a dimple in a nuclear fuel spacer grid. First, the strains of curved part of dimple are characterized via a comparison with strain solutions in pure bending. A reference 2D finite element (FE) model of 1-step stamping is then established, and the corresponding maximum strain is obtained. By varying the values of design variables of the 1st stamping tool in the 2-step stamping model, FE solutions are obtained to express the strain as a function of process variables, which provides the optimum values of process variables. Finally, applying these optimum values to a 3D FE model, we demonstrate the enhanced formability of the proposed 2-step stamping model.

• Transactions of the Korean Society of Mechanical Engineers B
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• v.40 no.5
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• pp.329-337
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• 2016

As a turbulence-enhancing device, a mixing vane, which is installed at a spacer grid of the fuel assembly, plays an important role in improving convective heat transfer by generating either swirl flow in the subchannels or cross flow between the fuel rod gaps. Therefore, both the geometric configuration and the arrangement pattern of a mixing vane are important factors in determining the performance of a mixing vane. In this study, in order to examine the flow-distribution features inside a $5{\times}5$ fuel assembly with split-type mixing vanes, which was used in the benchmark calculation of the OECD/NEA, we conduct simulations using the commercial computational fluid dynamics software, ANSYS CFX R.14. We compare the predicted results with measured data obtained from the MATiS-H (Measurement and Analysis of Turbulent Mixing in Subchannels-Horizontal) test facility. In addition, we discuss the effect of the split-type mixing vanes on the flow pattern inside the fuel assembly.

• Bulletin of the Korean Chemical Society
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• v.23 no.12
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• pp.1744-1748
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• 2002

Trifluoroacetate anion as a connector has been studied on $AgCF_3CO_2$ with 3,3'-$Py_2X$(X=O vs S) produces 1 : 1 adducts of [Ag($CF_3CO_2$)(3,3'-$Py_2X<$)]. Crystallographic characterization of [Ag($CF_3CO_2$)(3,3'-$Py_2X$)](monoclinic $P2_1$a=7.383(1)$\AA$b=19.801(3)$\AA$c=9.297(3)$\AA$,$\beta$=$100.26(2)^{\circ}$,V=1337.4(5) $\AA^3$, Z=2, R=0.0386) reveals that the 3,3'-$Py_2O$ spacer connects two silver ions to give a single strand and that the single strands are linked via the trifluoroacetate anions in an "up and down even-bridge" to give an elegant molecular grid. The framework of [$Ag(CF_3CO_2)(3,3'-Py_2X)$](monoclinic $P2_1/c$a=8.331(2)$\AA$b=14.010(2)$\AA$,c=11.926(3 $\AA$$\beta$=$93.70(2)^{\circ}$=1385.1(6)$\AA^3$, Z=4, R=0.0589) is a single-strand. The single strands are connected via the trifluoroacetate anions in a double-bridge, resulting in a typical molecular chicken-wire. The trifluoroacetate anion as a connector appears to be primarily associated with its moderately coordinating ability. Their structural features have been discussed based on the anion exchangeability. Thermal analyses indicate that the compounds are stable up to approximately $200^{\circ}C$.

• Nuclear Engineering and Technology
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• v.25 no.4
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• pp.552-560
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• 1993

In this study, the detailed fuel assembly stress analysis model to evaluate the structural integrity for seismic and blowdown accidents is developed. For this purpose, as the first step, the program MAIN which identifies the worst bending mode shaped fuel assembly(FA) in core model is made. And the finite element model for stress calculation of FA components is developed. In the model the fuel rods (FRs) and the guide thimbles are modelled by 3-dimensional beam elements, and the spacer grid spring is modelled by a linear and relational spring. The constraints come from the results of the program MAIN. The stress analysis of the 16$\times$16 type FA under arbitary seismic load is performed using the developed program and modelling technique as an example. The developed stress model is helpful for the stress calculation of FA components for seismic and blowdown loads to evaluate the structural integrity of FA.

• Transactions of the Korean Society of Mechanical Engineers A
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• v.37 no.1
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• pp.67-75
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• 2013

The flow stress of zircaloy-4 used in the spacer grid supporting a nuclear fuel rod was determined by the Johnson-Cook model, and model parameters were determined using reverse engineering. Parameters such as A, B, n and $\dot{\varepsilon}_0$ were determined by the tensile test result. To obtain the parameters C and m, a slot milling test and numerical simulation were performed. The objective functions were defined as the difference between the experimental and the simulation results, and then, the parameters were determined by minimizing the objective function. To verify the validity of the determined parameters, cross-verification for each case was conducted through a shearing test and simulation. The results tend to show agreement with the experimental results, such as the features of sheared edges and maximum punch force, with the correlation coefficients exceeding at least 0.97.

• Transactions of the Korean Society of Mechanical Engineers B
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• v.40 no.12
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• pp.815-824
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• 2016

The fuel assembly for pressurized water reactor (PWR) consists of fuel rod bundle, spacer grid and bottom/top end fittings. The cooling water in high pressure and temperature is introduced in lower plenum of reactor core and directed to upper plenum through the subchannel which is formed between the fuel rods. The main thermal-hydraulic performance parameters for the PWR fuel are pressure drop and critical heat flux in normal operating condition, and quenching time in accident condition. The Korea Atomic Energy Research Institute (KAERI) has been developing an advanced PWR fuel, dual-cooled annular fuel and accident tolerant fuel for the enhancement of fuel performance and the localization. For the key thermal-hydraulic technology development of PWR fuel, the KAERI LWR fuel team has conducted the experiments for pressure drop, turbulent flow mixing and heat transfer, critical heat flux(CHF) and quenching. The computational fluid dynamics (CFD) analysis was also performed to predict flow and heat transfer in fuel assembly including the spent fuel assembly in dry cask for interim repository. In addition, the research cooperation with university and nuclear fuel company was also carried out to develop a basic thermal-hydraulic technology and the commercialization.