• Proceedings of the Korean Society for Noise and Vibration Engineering Conference
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• 2010.05a
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• pp.268-269
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• 2010

• Nuclear Engineering and Technology
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• v.26 no.3
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• pp.440-452
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• 1994

Nuclear fuel bundles are designed such that the heat flux at a-fuel pin surface should not exceed the critical heat flux (CHF) during normal operation and anticipated transient. Therefore, evaluation of the CHF for fuel bundle is demanded in an exact and reliable manner. One of the major concerns with the current application of CHF correlations is that the CHF based on circular tubes is applied to the fuel bundle subchannel analysis, mainly in terms of the hydraulic diameter with correction factors which may result in a source of possibly large uncertainties in CHF prediction. The hydraulic diameter does not recognize the local properties of fluid nor such effect as the surface curvature; the turbulence action on the convex surface is much more pronounced than that on the concave surface. Even for the tube having concave curvature, the effect of tube diameter on CHF becomes important with decreasing diameter. These facts imply that the convex curvature effect is significant and crucial to the reliable CHF prediction. This paper reviews and discusses analytical and experimental aspects of effect of transverse convex curvature in incompressible turbulent flow and heat transfer, and on CHF. Flow models to quantify this effect are briefly mentioned and future works are recommended.

• Nuclear Engineering and Technology
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• v.25 no.1
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• pp.166-177
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• 1993

A simplified thermal analysis method to evaluate the maximum temperature of the CANDU 37-element fuel bundle within a fuel basket in a given spent fuel dry storage canister has been presented along with the results of sample analyses performed to examine the parametric effects of the ambient conditions on the maximum fuel temperature within a canister. To solve the multi-dimensional heat transfer problem of the complex geometry of rod bundles within a canister where three modes of heat transfer are superimposed, the CANDU spent fuel bundles stored in the dry storage canister are first replaced by equivalent concentric fuel cylinders. The simplified axi-symmetric two-dimensional multi-mode heat transfer problem of the equivalent fuel cylinders is then analyzed with an existing computer code, HEATING5, using additional input data and heat transfer correlations. A comparison between the predicted temperature profile and the mock-up test results shows that the agreement is quite satisfactory.

• Nuclear Engineering and Technology
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• v.2 no.1
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• pp.19-25
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• 1970

A study was conducted on the coolant mixing between water flowing in two adjacent subchannels. Measurements were made of the quantity of mass transferred between a larger rectangular channel and a smaller triangular channel in a 19-rod fuel bundle under the conditions of single phase flow and air-water two-phase flow. The results of the experiments showed that the low mixing rate appears in single phase flow, and high mixing rate was measured in air-water two-phase flow Mixing rate decreases with the increasing of air void fraction during the air-water flow. It seems that the high mixing rate in the air-water flow was caused due to adequate agitation of the chaotic air void.

• Current Industrial and Technological Trends in Aerospace
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• v.6 no.2
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• pp.80-89
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• 2008

Photogrammetry is a non-direct 3-dimensional coordinate measurement technique using 2-dimensional photographic images. For reconstruction 3-dimensional data from the 2-dimensional photos, photogrammetry uses the fundamental principle of triangulation. Digital photogrammetry solve for the camera location and coordinates simultaneously through the mapping, scaling and bundle processing in software processing. In this paper, several applications for photogrammetry measurement are introduced, such as 'photogrammetric measurement of the gravity deformation of a cassegrain type antenna', 'analysis of photogrammetry data from ISIM mockup', 'underwater photogrammetric verification of nuclear fuel assemblies', 'spacecraft optical bench measurement' and 'spacecraft ground support equipment measurement'.

• The KSFM Journal of Fluid Machinery
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• v.17 no.6
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• pp.89-94
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• 2014

The 4th generation SFR is being designed with a milestone of construction by 2028. It is important to understand the subchannel flow characteristics in fuel assembly through the experimental investigations and to estimate the calculation uncertainties for insuring the confidence of the design code calculation results. The friction coefficient and the mixing coefficient are selected as primary parameters. The two parameters are related to the flow distribution and diffusion. To identify the flow distribution, an iso-kinetic method was developed based on the previous study. For the mixing parameters, a wire mesh system and a laser induced fluorescence methods were developed in parallel. The measuring systems were adopted on 37 rod bundle test geometry, which was developed based on the Euler number scaling. A scaling method for a design of experimental facility and the experimental identification techniques for the flow distribution and mixing parameters were developed based on the measurement requirement.

• Nuclear Engineering and Technology
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• v.52 no.9
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• pp.1945-1954
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• 2020

Subchannel code is one of the effective simulation tools for thermal-hydraulic analysis in nuclear reactor core. In order to reduce the computational cost and improve the calculation efficiency, empirical correlation of turbulent mixing coefficient is employed to calculate the lateral mixing velocity between adjacent subchannels. However, correlations utilized currently are often fitted from data achieved in central channel of fuel assembly, which would simply neglect the wall effects. In this paper, the CFD approach based on spectral element method is employed to predict turbulent mixing phenomena through gaps in 3 × 3 bare tight lattice rod bundle and investigate the flow pulsation through gaps in different positions. Re = 5000,10000,20500 and P/D = 1.03 and 1.06 have been covered in the simulation cases. With a well verified mesh, lateral velocities at gap center between corner channel and wall channel (W-Co), wall channel and wall channel (W-W), wall channel and center channel (W-C) as well as center channel and center channel (C-C) are collected and compared with each other. The obvious turbulent mixing distributions are presented in the different channels of rod bundle. The peak frequency values at W-Co channel could have about 40%-50% reduction comparing with the C-C channel value and the turbulent mixing coefficient β could decrease around 25%. corrections for β should be performed in subchannel code at wall channel and corner channel for a reasonable prediction result. A preliminary analysis on fluctuation at channel gap has also performed. Eddy cascade should be considered carefully in detailed analysis for fluctuating in rod bundle.

• Nuclear Engineering and Technology
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• v.54 no.7
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• pp.2714-2719
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• 2022

The present paper proposes some new computational methods and results in the framework of flow computation through congested domains seen as porous media, as it can be found in the core of a Pressurized Water Reactor (PWR). The flow is thus mostly governed by the distribution of pressure losses, both through the porous structures, such as fuel assemblies, and in the thin fluid layers between them. The purpose of the present paper is to consider the question of the interaction of a flow and a rod bundle from an analytical point of view gathering all the contributions through a set of equations as simple and representative as possible. It aims at demonstrating a sound understanding of the relevant phenomena governing the flow establishment in the geometry of interest instead of relying mainly on a posteriori observations obtained both experimentally and numerically. Comparison with two set of experimental results showed good agreement. The model proposed being analytical it appears easily implementable for studies needing an expression of fluid forces in a rod array as for fuel assembly bowing issue. It would be interesting to test the reliability of the model on other geometry with different P/R ratios.

• Transactions of the Korean Society of Pressure Vessels and Piping
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• v.19 no.1
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• pp.27-35
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• 2023

Pressure tubes are the main components of PHWR core and serve as the pressure boundary of the primary heat transport system. However, because pressure tubes have changed their geometrical dimensions under the severe operating conditions of high temperature, high pressure and neutron irradiation according to the increase of operation time, all dimensional changes should be predicted to ensure that dimensions remain within the allowable design ranges during the operation. Among the deformations, the diameter expansion due to creep leads to the increase of bypass flow which may not contribute to the fuel cooling, the decrease of critical channel power and finally the deration of the power to maintain the operational safety margin. This study is focused on the modeling of the expansion of the pressure tube diameter based on the operating conditions and measured diameter data. The pressure tube diameter expansion was modeled using the neutron flux and temperature distributions of each fuel channel and each fuel bundle as well as the measured diameter data. Although the basic concept of the current modeling approach is simple, the diameter prediction results using the developed methodology showed very good agreement with the real data, compared to the existing methodology.

• 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.