• Nuclear Engineering and Technology
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• v.54 no.11
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• pp.4296-4309
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• 2022

Mesh and turbulence model sensitivity analyses have been performed on computational fluid dynamics simulations executed with Hydra and ANSYS Fluent for a single CANadian Deuterium Uranium (CANDU) 37M nuclear fuel bundle placed within a standard pressure tube. The goal of this work was to perform a methodical analysis to objectively determine an appropriate mesh and to gauge the sensitivity of different turbulence models for CANDU subchannel flow under isothermal conditions. The boundary conditions and material properties are representative of normal operating conditions in a high-powered channel of the Darlington Nuclear Generating Station. Four meshes were generated with ANSYS Workbench Meshing, ranging from 22 to 84 million cells, and analyzed here to determine an appropriate level of mesh resolution and quality. Five turbulence models were compared in the turbulence model sensitivity analysis: standard k - ε, RNG k - ε, realizable k - ε, SST k - ω, and the Reynolds Stress Model. The intent of this work was to gain confidence in mesh generation and turbulence model selection of a single bundle to inform the decision making of subsequent investigations of an entire fuel channel containing a string of twelve bundles.

• The Journal of the Acoustical Society of Korea
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• v.6 no.2
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• pp.19-29
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• 1987

The absorption coefficients of various length bundles of straws simulating perforated material were studied both theoretically and experimentally. For the theoretical predictions Zwikker and Kosten's theory was modified by adapting Biot's theory based on Poiseuille flow. The experimental data were collected using an impedance tube where the attenuation along the length of the tube was considered. The theoretically predicted values agreed very well with the experimentally measured ones for frequencies lower than 700Hz with bundles shorter than 120mm in length placed against the rigid end of the impedance tube. Configurations with an air gap between the end of a bundle and the rigid end were also investigated. Absorption coefficients were higher for 150mm bundles than for those of combined/air gap configurations with a total length of 150mm. Also for the fixed bundle lengths, absorption was found to increase with increasing air gap.

• Journal of Mechanical Science and Technology
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• v.17 no.7
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• pp.947-957
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• 2003

The pressure tube is a major component of the CANDU reactor, which supports nuclear fuel bundle. In order to complete the integrity evaluation of pressure tube, expert knowledge, iterative calculation procedures and a lot of input data are required. More over, results of integrity assessment may be different according to the evaluation method. For this reason, an integrity evaluation system, which provides efficient way of evaluation with the help of attached database, was developed. The present system was built on the basis of 3D FEM results, ASME Sec. XI, and Fitness For Service Guidelines for CANDU pressure tubes issued by the AECL (Atomic Energy Canada Limited). The present system also covers the delayed hydride cracking and the blister evaluation, which are the characteristics of pressure tube integrity evaluation. In order to verify the present system, several case studies have been performed and the results were compared with those from AECL. A good agreement was observed between those two results.

• Journal of Advanced Marine Engineering and Technology
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• v.34 no.4
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• pp.491-499
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• 2010

Heat transfer performance of tube bundles have long been investigated since they were widely used. Most of previous experimental and numerical works for tube bundles were performed with tube diameter in the range of 25~51mm and Reynolds number of $8.000{\leq}Re{\leq}30.000$. Recently, tube bundles with small diameter tube collects interests since the mini-channel tube provides higher compactness. The present work aims to investigate the applicability of previous correlations available in the open literature to the tube bundles with small diameter of 1.5mm and $3.000{\leq}Re{\leq}7.000$. A commercial CFD package was used to analyze the thermal-hydraulic performance of them. The results show that the Zukauskas correlation developed for larger diameter tube and higher Reynolds number are still in good agreement with them within the discrepancy of 4.7%. The analyses also show that the Nuselt number increases with a decrease in the longitudinal pitch.

• Journal of Advanced Marine Engineering and Technology
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• v.36 no.4
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• pp.497-503
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• 2012

The aim of this study was to optimize S-tube shape of heat exchanger in term of reducing the size of tube bundle and improving the mechanical properties such as the thermal stress and resonance. The geometric parameters such as offset length, the straight distance between one end and other end of tube, the tube length in straight portion and fillet radius was assessed as a valid parameters. The structural analysis was performed to estimate the structural characteristics. Main effect analysis was performed to investigate the main effect for the various geometric parameters. The response surface methodology was employed to establish mathematical approximation models as a function of the geometric parameters of the S-tube. Also, The optimization was performed to optimize geometric parameters of S-tube using the regression equations and optimization tool. The optimized tube shape has been proposed. Those could be used in the heat exchanger design used in high temperature.

• Nuclear Engineering and Technology
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• v.25 no.3
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• pp.381-393
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• 1993

This paper presents the experimental results on flooding of countercurrent flow in vertical multirod channels, which consists of falling water film and upward air flow. In particular, the effects of spacer grids, with and without mixing vane, and of blockage in the multirod bundle on the behaviour of flooding were investigated. The 5$\times$5 zircaloy tube bundle was used for the test section. The comparison of previous analytical models and empirical correlations with present data on flooding showed that the existing models and correlations predict much higher flooding curves. The spacer grid causes the lower flooding air flow rate to compare with the bare rod bundle. However, the mixing spacer grids need a higher flooding air flow rate for a constant liquid flow rate than the spacer grids without mixing vanes. The bundle containing blockages has the highest flooding air flow rate among the bundles with spacer grids and blockages. Empirical flooding correlations for the three types of test section have been made.

• KSTLE International Journal
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• v.10 no.1_2
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• pp.33-36
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• 2009

The nuclear steam generator is composed of a bundle of tubes. The length of these tubes is very long, but their diameter is small. Fluid exists inside of the steam generator and its flow causes vibration, therefore these tubes are supported by anti-vibration bars. The wear damage due to the vibration is known as fretting wear, which should be minimized to ensure the safety of the plants. Research needs to be done about decreasing the amount of fretting wear. Hard coatings have proven to be very effective in reducing the amount of wear. The commercial coatings of TiN and CrN have excellent wear resistance and are used to protect the Inconel tube from fretting wear. The tube-on-flat type tester was used for conducting the fretting wear tests. It was found that the wear amounts of the coated tubes decreased depending on the coating thickness. CrN was found to be very effective in reducing the wear, while the wear amounts were dependent on the coating thickness in the case of TiN and a thick coating of TiN was very effective on wear resistance.

• Proceedings of the KSME Conference
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• 2000.11a
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• pp.843-848
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• 2000

The pressure tube is a major component of the CANDU reactor, which supports nuclear fuel bundle and it's containment vessel. If a flaw is found during the periodic inspection from the pressure tubes, the integrity evaluation must be carried out, and the safety requirements must be satisfied for continued service. In order to complete the integrity evaluation, complicated and iterative calculation procedures are required. Besides, a lot of data and knowledge for the evaluation are required for the entire integrity evaluation process. For this reason, an integrity evaluation system, which provides efficient way of evaluation with the help of attached databases, was developed. The developed system was built on the basis of ASME Sec. XI and FFSG(Fitness For Service Guidelines for zirconium alloy pressure tubes in operating CANDU reactors) issued by the AECL, and covers the delayed hydride cracking(DHC). Various analysis methods are provided for the integrity evaluation of pressure tube. In order to verify the developed system, several case studies have been performed and the results were compared with those from AECL. A good agreement was observed between those two results.

• 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.41 no.11
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• pp.719-726
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• 2017

An experimental study was performed to investigate the combined effects of the pitch and heat flux of nearby tubes on boiling in a pool as well as the heat transfer from a horizontally-installed tube bundle. For this test, two smooth stainless steel tubes (19 mm outside diameter) were used, and the water was at atmospheric pressure. The pitch of these tubes was varied between 28.5 mm and 95 mm, and the heat flux of the nearby tube altered between 0 and $90kW/m^2$. Enhancements in heat transfer were clearly observed when the heat flux of the nearby tube increased while the heat flux of the test section remained below $40kW/m^2$. The tube pitch was found to have a negligible effect on heat transfer when the pitch was greater than four times larger than the tube diameter. The circulating flow, convective flow, and liquid agitation were all seen to enhance heat transfer; however, the interaction between the flow and coalescence of bubbles was detrimental to heat transfer.