• Proceedings of the Korean Nuclear Society Conference
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• 2002.10a
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• pp.145-145
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• 2002

• Korean Journal of Air-Conditioning and Refrigeration Engineering
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• v.29 no.10
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• pp.538-550
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• 2017

In a PWR (Pressurized Water Reactor), the appropriate heat removal from the surface of fuel rod bundle is important for ensuring thermal margins and safety. Although many CFD (Computational Fluid Dynamics) software have been used to predict complex flows inside fuel assemblies with mixing vanes, there is no domestic regulatory guideline for the comprehensive evaluation of CFD software. Therefore, from the nuclear regulatory perspective, it is necessary to perform the systematic assessment and prepare the domestic regulatory guideline for checking whether valid CFD software is used for nuclear safety problems. In this study, to provide systematic evaluation and guidance on the applicability of CFD software to the domestic nuclear safety area, the results of the sensitivity analysis for the effect of the discretization scheme accuracy for the convection terms and turbulence models, which are main factors that contribute to the uncertainty in the calculation of the nuclear safety problems, on the prediction performance for the turbulent flow distribution inside the fuel assembly with split-type mixing vanes were explained.

• Journal of Mechanical Science and Technology
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• v.15 no.3
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• pp.403-409
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• 2001

The main objective of the present study is to perform an experimental evaluation of five existing correlations for the subchannel pressure drop analysis of a wire-wrapped fuel assembly. For this purpose, a series of water experiments have been performed using a helical wire-wrapped 19-pin fuel assembly for various test parameters. Four different test sections with different pitch to rod diameter ratios (P/D) and wire lead length to rod diameter ratios (H/D) have been fabricated. A series of pressure drop measurements were made to obtain friction factors for these four test sections. The new data along with existing data are used to evaluate existing correlations. Both the original and the simplified Cheng and Todreas correlations give the best agreement with experimental data for all flow regions.

• Proceedings of the Korean Radioactive Waste Society Conference
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• 2017.10a
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• pp.91-92
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• 2017

In this study, assembling procedure of TRU fuel assembly was reviewed and divided into rod bundle assembling, mating preassemblies and welding, and inspection and non-destructive examination. Based on this assumption, the design criteria of a remote assembler for TRU fuel assembly of PGSFR is defined and predictable technical issues are proposed.

• Transactions of the Korean Society of Mechanical Engineers
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• v.18 no.11
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• pp.2986-2998
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• 1994

• Proceedings of the SAREK Conference
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• 2004.11a
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• pp.126-126
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• 2004

• Proceedings of the SAREK Conference
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• 2005.06a
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• pp.131-131
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• 2005

• Transactions of the Korean Society of Mechanical Engineers B
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• v.41 no.1
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• pp.53-62
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• 2017

The forced convection heat transfer performance of a twist-vane spacer grid for a dual-cooled annular fuel assembly was examined experimentally. The twist-vane spacer grid was uniquely designed to enhance mixing inside subchannels and mixing between adjacent subchannels. For testing, a $4{\times}4$ square-arrayed rod bundle with narrow gaps between rods was prepared as the dual-cooled annular fuel assembly to be simulated. The pitch-to-rod diameter ratio of simulated dual-cooled annular fuel assembly was 1.08. The experiments were performed under the following conditions: axial bulk velocity, 1.5 m/s and heat flux, $26kW/m^2$. With regard to the circumferential temperature distribution, the lowest rod-wall temperatures upstream and downstream were measured at the subchannel center and the position toward the tip of twist-vane, respectively. With regard to the axial temperature distribution, behind the twist-vane spacer grid, the rod-wall temperature decreased drastically, and the Nusselt number was enhanced by up to 56 %. The present measured data indicate that the twist-vane spacer grid can effectively improve the forced convection heat transfer in the dual-cooled annular fuel assembly with narrow gaps.

• Nuclear Engineering and Technology
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• v.55 no.12
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• pp.4597-4606
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• 2023

Thermal analysis of the CANDU spent fuel dry storage canister is very important to ensure the integrity of the spent fuel. The analyses have been conducted using a conservative approach, with a particular focus on the peak cladding temperature (PCT) of the fuel rods in the canister. In this study, we have performed a realistic thermal analysis using a computational fluid dynamics (CFD) code. The canister contains 9 fuel bundle baskets. A detailed analysis of even a single basket requires significant computational resources. To overcome this challenge, we replaced each basket with an equivalent heat conductor (EHC), of which effective thermal conductivity (ETC) is developed from the results of detailed CFD calculations of a fuel bundle basket. Then, we investigated the effects of some conservative models, ultimately aiming at a realistic analysis. The results revealed: (i) The influence of convective heat transfer in the basket cannot be ignored, but it's less significant than expected. (ii) Modeling of the lifting rod is crucial, as it plays a decisive role in axial heat transfer at the center of the canister and significantly reduces the PCT. (iii) Convection within the canister is very important, as it not only reduces the PCT but also shifts its location upwards.

• Nuclear Engineering and Technology
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• v.30 no.5
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• pp.452-461
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• 1998

The pin power distribution is determined by analyzing the rotational gamma scanning data for 36 element fuel bundle of HANARO. A fission monitor of Nb$^{95}$ is chosen by considering the criteria of the half-life, fission yield, emitting ${\gamma}$-ray energy and probability. The ${\gamma}$-ray spectra were measured in Korea Atomic Energy Research Institute(KAERI) by using a HPGe detector and by rotating the fuel bundle at steps of 10$^{\circ}$. The counting rates of Nb$^{95}$ 766 keV ${\gamma}$-rays are determined by analyzing the full absorption peak in the spectra. A 36$\times$36 response matrix is obtained from calculating the contribution of each rod at every scanning angle by assuming 2-dimensional and parallel beam approximations for the measuring geometry. In terms of the measured counting rates and the calculated response matrix, an inverse problem is set up for the unknown distribution of activity concentrations of pins. To select a suitable solving method, the performances of three direct methods and the iterative least-square method are tested by solving simulation examples. The final solution is obtained by using the iterative least-square method that shows a good stability. The influences of detection error, step size of rotation and the collimator width are discussed on the accuracy of the numerical solution. Hence an improvement in the accuracy of the solution is proposed by reducing the collimator width of the scanning arrangement.