• Nuclear Engineering and Technology
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• v.47 no.5
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• pp.523-533
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• 2015

Three-dimensional flow phenomena in a wire-wrapped 37-pin fuel assembly mock-up of a Japanese loop-type sodium-cooled fast reactor, Monju, were investigated with a numerical analysis using a general-purpose commercial computational fluid dynamics code, CFX. Complicated and vortical flow phenomena in the wire-wrapped 37-pin fuel assembly were captured by a Reynolds-averaged Navier-Stokes flow simulation using a shear stress transport turbulence model. The main purpose of the current study is to understand the three-dimensional complex flow phenomena in a wire-wrapped fuel assembly to support the license issue for the core design. Computational fluid dynamics results show good agreement with friction factor correlation models. The secondary flow in the corner and edge subchannels is much stronger than that in an interior subchannel. The axial velocity averaged in the corner and edge subchannels is higher than that averaged in the interior subchannels. Three-dimensional multiscale vortex structures start to be formed by an interaction between secondary flows around each wire-wrapped pin. Behavior of the large-scale vortex structures in the corner and edge subchannels is closely related to the relative position between the hexagonal duct wall and the helically wrapped wire spacer. The small-scale vortex is axially developed in the interior subchannels. Furthermore, a driving force on each wire spacer surface is closely related to the relative position between the hexagonal duct wall and the wire spacer.

• 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.

• Nuclear Engineering and Technology
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• v.51 no.3
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• pp.665-675
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• 2019

The wire-wrapped fuel bundle is an assembly design in a sodium-cooled fast reactor. A wire spacer is used to maintain a constant gap between rods and to enhance the mixing of coolants. The wire makes the flow complicated by creating a sweeping flow and vortex flow. The vortex affects the flow field and heat transfer inside the subchannels. However, studies on vortices in this geometry are limited. The purpose of this research is to investigate the vortex flow created in the wire-wrapped fuel bundle. For analysis, a RANS-based numerical analysis was conducted for a 37-pin geometry. The sensitivity study shows that simulation with the shear stress transport model is appropriate. For the case of Re of 37,100, the mechanisms of onset, periodicity, and rotational direction were analyzed. The vortex structures were reconstructed in a three-dimensional space. Vortices were periodically created in the interior subchannel three times for one wire rotation. In the edge subchannel, the largest vortex occurred. This large vortex structure blocked the swirl flow in the peripheral region. The small vortex formed in the corner subchannel was negligible. The results can help in understanding the flow field inside subchannels with sweeping flow and vortex structures.

• Transactions of the Korean Society of Mechanical Engineers
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• v.18 no.5
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• pp.1235-1244
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• 1994

In this study, experimental results of the nucleate boling of wire-wrapped tubes are provided. Both water and R-113 were boiled. Solid stainless steel wires, stranded copper wires and stranded nylon wires were tested. Solid stainless wire is effective to enhance the boiling of water. The performance is approximately the same(or slightly better at certain conditions) as that of GEWA-T tubes. For the test range of wire diameter 0.6 $mm{\le}d{\le}2.6 mm$, the optimum gap width increases as the wire diameter increases. The maximum heat transfer coefficient was obtained for the 1.0 mm diameter wire, and it is 1.6 times larger than that of the GEWA-T at the heat flux of 20 $kW/m^{2}.$ Solid stainless wire is also effective to enhance the boling of R-113 at low heat fluxes. The performance of the wire-wrapped tube approaches that of GEWA-T. At high heat fluxes, however, the enhancement decreases. The reason may be attributed to the cavity shape and the high wettability of the refrigerants. Stranded copper or nylon wire is effective to enhance the boiling of R-113. The performance is approximately the same(or slightly better) as that of GEWA-T tubes. Maximum heat transfer was obtained for the stranded nylon wire, and it is approximately 1.4 times larger than that for the GEWA-T at the heat flux of 20 $kW/m^{2}.$ The reason may be atrributed to the favorable thermal environment in the restricted regions formed by twisted wires.

• Nuclear Engineering and Technology
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• v.56 no.4
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• pp.1125-1134
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• 2024

This paper presents a validation study of the subchannel analysis code SLTHEN used for the core thermal-hydraulic design of the Prototype Gen-IV sodium-cooled fast reactor (PGSFR). To assess the performance of the ENERGY model of SLTHEN, four liquid metal heating experiments conducted by ORNL, WARD, and KIT with hexagonal assemblies of wire-wrapped rod bundles were analyzed. These experiments were performed with 19-and 61-pin bundles and varying power distributions of axial and radial peaking factors up to 1.4 and 3.0, respectively. The coolant subchannel temperatures measured at different axial locations were compared with the SLTHEN predictions with the Novendstern, Chiu-Rohsenow-Todreas (CRT), and Cheng-Todreas (CT) correlations for flow split and mixing in wire-wrapped pin bundles. The results showed that the SLTHEN predicts the measured subchannel temperatures reasonably well with root-mean-square errors of ~10 % and maximum errors of ~20 %. It was also observed that the CRT and CT correlations consistently outperform the Novendstern correlation.

• Journal of Mechanical Science and Technology
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• v.20 no.9
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• pp.1514-1524
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• 2006

This work presents the three-dimensional analysis of flow and heat transfer performed for a wire-wrapped fuel assembly of liquid metal reactor using Reynolds-averaged Wavier-Stokes analysis in conjunction with 557 model as a turbulence closure. The whole fuel assembly has been analyzed for one period of the wire-spacer using periodic boundary conditions at inlet and outlet of the calculation domain. Three different assemblies, two 7-pin wire-spacer fuel assemblies and one bare rod bundle, apart from the pressure drop calculations for a 19-pin case, have been analyzed. Individual as well as a comparative analysis of the flow field and heat transfer have been discussed. Also, discussed is the position of hot spots observed in the wire-spacer fuel assembly. The flow field in the subchannels of a bare rod bundle and a wire-spacer fuel assembly is found to be different. A directional temperature gradient is found to exist in the subchannels of a wire-spacer fuel assembly Local Nusselt number in the subchannels of wire-spacer fuel assemblies is found to vary according to the wire-wrap position while in case of bare rod bundle, it's found to be constant.

• Nuclear Engineering and Technology
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• v.48 no.2
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• pp.376-385
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• 2016

A hexagonally arrayed 37-pin wire-wrapped rod bundle has been chosen to provide the experimental data of the pressure loss and flow rate in subchannels for validating subchannel analysis codes for the sodium-cooled fast reactor core thermal/hydraulic design. The iso-kinetic sampling method has been adopted to measure the flow rate at subchannels, and newly designed sampling probes which preserve the flow area of subchannels have been devised. Experimental tests have been performed at 20-115% of the nominal flow rate and $60^{\circ}C$ (equivalent to Re ~ 37,100) at the inlet of the test rig. The pressure loss data in three measured subchannels were almost identical regardless of the subchannel locations. The flow rate at each type of subchannel was identified and the flow split factors were evaluated from the measured data. The predicted correlations and the computational fluid dynamics results agreed reasonably with the experimental data.

• Proceedings of the Korean Nuclear Society Conference
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• 2001.10a
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• pp.197.1-197
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• 2001

• Journal of Applied Reliability
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• v.5 no.4
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• pp.405-412
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• 2005

In this paper, we verified the failure mechanism of the discolored masking tape wrapped around the wire and studied an improved case study as well. Masking tapes are poly(Vinylchloride-co-Methyl methacrylate) (PVC) and the wire connecting with indoor/outdoor components of the air conditioner is chloroprene(CR). The masking tape under. the ultraviolet ray was accelerated to discolor fivefold. The results determined reveal the discoloration of masking tapes as an elution of the additives, dephenyl Amine series antioxidants, in the connecting wire. When changed dephenyl Amine series to phenol series, we found that the degree of discoloration was dramatically decreased. Due to stink of Thiourea and Thiuram monosulfide, as accelerators in CR wire, we controlled the vulcanization temperature and time without using these accelerators and could get the wire not to smell bad with the equal mechanical properties.

• Nuclear Engineering and Technology
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• v.36 no.2
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• pp.165-174
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• 2004

The distributed resistance model has been recently implemented into the MATRA-LMR code in order to improve its prediction capability over the wire-wrap model for a flow blockage analysis in the LMR. The code capability has been investigated using experimental data observed in the FFM (Fuel Failure Mock-up)-2A and 5B for two typical flow conditions in a blocked channel. The predicted results by the MATRA-LMR with a distributed resistance model agreed well with the experimental data for wire-wrapped subchannels. However, it is suggested that the parameter n in the distributed resistance model needs to be calibrated accurately for a reasonable prediction of the temperature field under a low flow condition. Finally, the analyses of a blockage for the assembly of the KALIMER design are performed. Satisfactory results by the MATRA-LMR code were obtained through and rerified a comparison with results of the SABRE code.