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Journal articles on the topic 'Wire-wrapped rod bundles'

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Published: 10 December 2022

Last updated: 28 January 2023

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1

Ninokata, Hisashi, Apostolos Efthimiadis, and Neil E. Todreas. "Distributed resistance modeling of wire-wrapped rod bundles." Nuclear Engineering and Design 104, no. 1 (October 1987): 93–102. http://dx.doi.org/10.1016/0029-5493(87)90306-2.

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2

Suh, K. Y., N. E. Todreas, and W. M. Rohsenow. "Mixed Convective Low Flow Pressure Drop in Vertical Rod Assemblies: II—Experimental Validation." Journal of Heat Transfer 111, no. 4 (November 1, 1989): 966–73. http://dx.doi.org/10.1115/1.3250812.

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An experimental study has been conducted to validate the predictive models and correlations for laminar and transition flow frictional pressure loss in vertical rod bundles under mixed convection conditions. An experimental procedure has been developed to measure low differential pressures under mixed convection conditions in 19 heated rod bare and wire-wrapped assemblies. The proposed model has been found successfully to predict the effects of wire wrapping, power skew, transition from laminar regime, developing and Interacting flow redistributions, and rod number on the friction loss characteristics in bundle geometries over the bundle average Grq/Re number range of 6 to 18,500.

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3

Dix, Adam, and Seungjin Kim. "A novel friction factor model for wire-wrapped rod bundles." Nuclear Engineering and Design 401 (January 2023): 112104. http://dx.doi.org/10.1016/j.nucengdes.2022.112104.

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4

Bovati, Octavio, Mustafa Alper Yildiz, Yassin Hassan, and Rodolfo Vaghetto. "Pressure drop and flow characteristics in partially blocked wire wrapped rod bundles." Annals of Nuclear Energy 165 (January 2022): 108671. http://dx.doi.org/10.1016/j.anucene.2021.108671.

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5

Choi, Sun Rock, Hyungmo Kim, Seok-Kyu Chang, Hae Seob Choi, Dong-Jin Euh, Hyeong-Yeon Lee, and Won Sik Yang. "Assessment of subchannel flow mixing coefficients for wire-wrapped hexagonal fuel rod bundles." Annals of Nuclear Energy 166 (February 2022): 108810. http://dx.doi.org/10.1016/j.anucene.2021.108810.

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6

Bovati, Octavio, Mustafa Alper Yildiz, Yassin Hassan, and Rodolfo Vaghetto. "RANS simulations for transition and turbulent flow regimes in wire-wrapped rod bundles." International Journal of Heat and Fluid Flow 90 (August 2021): 108838. http://dx.doi.org/10.1016/j.ijheatfluidflow.2021.108838.

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7

Carajilescov, Pedro, and Elói Fernandez y Fernandez. "Model for subchannel friction factors and flow redistribution in wire-wrapped rod bundles." Journal of the Brazilian Society of Mechanical Sciences 21, no. 4 (December 1999): 589–99. http://dx.doi.org/10.1590/s0100-73861999000400003.

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8

Hu, Rui, and Thomas H. Fanning. "A momentum source model for wire-wrapped rod bundles—Concept, validation, and application." Nuclear Engineering and Design 262 (September 2013): 371–89. http://dx.doi.org/10.1016/j.nucengdes.2013.04.026.

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9

Kim, Hansol, Yu Min Chen, and Yassin Hassan. "Prediction of pressure drop in hexagonal wire-wrapped rod bundles using artificial neural network." Nuclear Engineering and Design 381 (September 2021): 111365. http://dx.doi.org/10.1016/j.nucengdes.2021.111365.

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10

Chen, S. K., Y. M. Chen, and N. E. Todreas. "The upgraded Cheng and Todreas correlation for pressure drop in hexagonal wire-wrapped rod bundles." Nuclear Engineering and Design 335 (August 2018): 356–73. http://dx.doi.org/10.1016/j.nucengdes.2018.05.010.

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11

Sorokin, A., Yu Kuzina, G. Sorokin, and N. Denisova. "MODELING OF HEAT AND MASS TRANSFER PROCESSES IN FUEL ASSEMBLIES OF FAST REACTORS AS PART OF THE CHANNEL-BY-CHANNEL CALCULATION METHOD. GENERALIZED EXCHANGE CHARACTERISTICS FOR SINGLE-PHASE FLOWS OF LIQUID METALS." PROBLEMS OF ATOMIC SCIENCE AND TECHNOLOGY. SERIES: NUCLEAR AND REACTOR CONSTANTS 2020, no. 2 (June 26, 2020): 104–30. http://dx.doi.org/10.55176/2414-1038-2020-2-104-130.

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The results of experimental and computational-theoretical studies of the transverse interchannel molecular-turbulent and convective exchange of mass, momentum and energy are presented in the framework of the channel-by-channel model of thermohydraulic calculation of fuel assemblies of fuel elements of the fast reactors core with liquid metal coolants for nominal, non-nominal and accident conditions. A complete system of inter-channel exchange coefficients is obtained for closing the system of macro-transport equations in a wide range of parameters, taking into account the fuel assemblies deformation. It was shown that transverse convective exchange in fuel assemblies with spacing by wire wrapped is caused by a deficit of static pressure in the region behind winding at the surface of the rod, and the high intensity of molecular-turbulent exchange in tight bundles is explained by the appearance of secondary flows in cells. The heat exchange between cells due to the thermal conductivity of the fuel rods is determined by the thermal modeling parameter of the fuel rods for the first harmonic. The effect of wire wrapped on the flow can be modeled by periodically changing hydraulic resistance to the transverse coolant flow. The integral model of convective inter-channel exchange in fuel assemblies with distance wire wrap is substantiated. Taking into account the centrifugal effect and calculating the interchannel exchange coefficients taking into account the local distribution of parameters allows us to clarify the thermohydraulic characteristics in the shaped bundles. For the characteristic parameters of fuel assemblies of fast reactors (Pe ≥ 50; S/d ≥ 1,1; h/d ≤ 30; ε1 < 1) the predominant is convective inter-channel exchange due to the spacing of fuel rods by wire wrapped. For highly heat-conducting fuel compositions, a significant contribution to the inter-channel heat exchange can be made by the exchange due to the thermal conductivity of the fuel elements. At a low flow rate (Pe <50), the contribution of molecular-turbulent exchange and heat exchange increases due to the thermal conductivity of the fuel rods.

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12

Cheng, Shih-Kuei, and Neil E. Todreas. "Hydrodynamic models and correlations for bare and wire-wrapped hexagonal rod bundles — Bundle friction factors, subchannel friction factors and mixing parameters." Nuclear Engineering and Design 92, no. 2 (April 1986): 227–51. http://dx.doi.org/10.1016/0029-5493(86)90249-9.

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13

Chen, Y. M., S. K. Chen, and Y. Hassan. "The blockage model of Upgraded Cheng and Todreas correlation for pressure drop in hexagonal wire-wrapped rod bundles." Nuclear Engineering and Design 395 (August 2022): 111874. http://dx.doi.org/10.1016/j.nucengdes.2022.111874.

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14

Bovati, Octavio, and Yassin Hassan. "Implementation of a CFD methodology for computing subchannel friction factors and split parameters in wire-wrapped rod bundles." Nuclear Engineering and Design 397 (October 2022): 111952. http://dx.doi.org/10.1016/j.nucengdes.2022.111952.

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15

Pacio, J., S. K. Chen, Y. M. Chen, and N. E. Todreas. "Analysis of pressure losses and flow distribution in wire-wrapped hexagonal rod bundles for licensing. Part II: Evaluation of public experimental data." Nuclear Engineering and Design 388 (March 2022): 111606. http://dx.doi.org/10.1016/j.nucengdes.2021.111606.

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16

Pacio, J., S. K. Chen, Y. M. Chen, and N. E. Todreas. "Analysis of pressure losses and flow distribution in wire-wrapped hexagonal rod bundles for licensing. Part I: The Pacio-Chen-Todreas Detailed model (PCTD)." Nuclear Engineering and Design 388 (March 2022): 111607. http://dx.doi.org/10.1016/j.nucengdes.2021.111607.

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17

Chen, S. K., Y. M. Chen, and N. E. Todreas. "Corrigendum to “The upgraded Cheng and Todreas correlation for pressure drop in hexagonal wire-wrapped rod bundles” [Nucl. Eng. Des. 335 (2018) 356–373]." Nuclear Engineering and Design 340 (December 2018): 414. http://dx.doi.org/10.1016/j.nucengdes.2018.10.009.

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18

Lyu, Kefeng, Xuelei Sheng, and Xudan Ma. "Thermal-hydraulic Assessment of Seven Wire-wrapped rod Bundle." E3S Web of Conferences 212 (2020): 01009. http://dx.doi.org/10.1051/e3sconf/202021201009.

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Lead bismuth eutectic (LBE) is one of the most potential materials for coolant and spallation target for Accelerator Driven Systems (ADS). Thermal-hydraulic behavior of LBE in fuel assembly is a key issue for development of the systems. To get a deeper understanding on the complex thermal-hydraulic features of wire-wrapped rod bundle cooled by upward LBE, an electrically bundle with 7 rods wrapped with helical wire was developed in KYLIN-II thermal-hydraulic forced circulation loop. The flow resistance, thermal entrance characteristic and heat transfer coefficient were investigated. As for the entrance characteristics, during the full heating length (exceeding 140 times the hydraulic diameter), the thermal field did not reach a fully developed and stable condition which is contrary to the ducted flows. The experimental heat transfer coefficient showed that the hexagonal shell has a great influence on the heat transfer coefficient in rod bundle geometry. For this reason the application of empirical correlation should be kept cautious in rod bundle analysis.

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19

LYU, Kefeng, Xuelei SHENG, Xudan MA, and Haitao WANG. "CFD Analysis of Scaled Wire-wrapped rod Bundle Cooled by LBE." E3S Web of Conferences 136 (2019): 01044. http://dx.doi.org/10.1051/e3sconf/201913601044.

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Lead bismuth eutectic (LBE) is one of the most potential materials for coolant for Lead based reactor and Accelerator Driven Systems (ADS). Thermal-hydraulic behaviour of LBE in fuel assembly is a key issue for development of the systems. To get a deeper understanding on the complex thermal-hydraulic features of wire-wrapped rod bundle cooled by upward LBE, CFD calculation based on RANS methodologies were also performed to support the experimental results analysis. The results concluded that LBE has the similar flow resistance characteristics with traditional fluids. Both the Rehme correlation and CFD showed a good agreement with the experimental results. As for the entrance characteristics, during the fully heating length (exceeding 140 times the hydraulic diameter), the thermal field did not reach a fully developed and stable condition which is contrary to the ducted flows. Based on the experimental results and CFD investigation of heat transfer coefficient showed that the hexagonal shell has a great influence on the heat transfer coefficient in rod bundle geometry. For this reason, the application of empirical correlation should be kept cautious in rod bundle analysis.

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20

Li, Yonghua, Meijun Li, and Yangyang Guo. "Analysis of Resistance Characteristics of a 37 Rod Fuel Bundle under Low Reynolds Number." Journal of Energy 2020 (December 2, 2020): 1–8. http://dx.doi.org/10.1155/2020/8861190.

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During the working period of decay heat removal system, the flow rate of liquid sodium in wire-wrapped fuel assembly is very low, generally Re < 1000 . In the present study, both experimental methods and numerical simulation methods are applied. First, water experiment of 37-pin wire-wrapped rod bundle was carried out. Then, the numerical simulation study was carried out, the experimental data and the numerical simulation results were compared and analyzed, and a suitable turbulence model was selected to simulate the liquid sodium medium. Finally, numerical simulations under different boundary conditions were performed. Results indicate that except for the low Reynolds number k - ε turbulence model, other turbulence models have little difference with the experimental results. The results of realizable k - ε turbulence model are the most close to the experimental results. Compared with the friction factor obtained by using water medium and liquid sodium medium, the calculation results of water medium and sodium medium under the same condition are basically consistent, with the deviation within 1%. The reason is that the velocity of water is higher than sodium medium at the same Reynolds number, and the transverse disturbance caused by helical wire is larger.

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21

Tan, Xuefeng, Bing Wang, Yun Guo, and Miao Hu. "Numerical Investigation of Special Heat Transfer Phenomenon in Wire-Wrapped Fuel Rod of SFR." Micromachines 13, no. 6 (June 11, 2022): 935. http://dx.doi.org/10.3390/mi13060935.

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Sodium-cooled reactors (SFR) have always been recognized as one of the most promising candidates for the fourth-generation nuclear systems as announced by the Generation-IV International Forum. In the design of SFR, helical wire-wrapped rod is applied to stabilize the structure of the rod bundle and enhance coolant mixing. Although there has been considerable research on SFR in computational fluid dynamics (CFD), the phenomenon of heat transfer has rarely been paid attention to. This article discovered that there exists reversed heat flux from coolant to wrapped wire, which is contrary to our usual understanding. This phenomenon has not been reported in previous CFD calculations. Hence, a solid heat conduction model is proposed to prove this phenomenon and analyze the heat transfer process. The simulation results show that the wrapping wire embedding depth, the shape of the calculation domain and the physical properties of all components have great influence on the magnitude of the reversed heat flux. The present findings will have strong influence on the temperature field and maximum value of the fuel rod as well as profound reference value for future flow calculation, especially in grid generation and treatment of the junction between the winding wire and fuel rod.

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22

Wang, Xi, and Xu Cheng. "Analysis of inter-channel sweeping flow in wire wrapped 19-rod bundle." Nuclear Engineering and Design 333 (July 2018): 115–21. http://dx.doi.org/10.1016/j.nucengdes.2018.04.008.

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23

Hou, Junren, Qingkang Song, Haojie Leng, Chaohui Xue, Yuan Yuan, and Yuan Zhou. "A non-destructive model for thermal-hydraulics of wire-wrapped rod bundle and wire-rod contact corner microscopic behavior." Progress in Nuclear Energy 154 (December 2022): 104469. http://dx.doi.org/10.1016/j.pnucene.2022.104469.

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24

Todreas, Neil, Shih-Kuei Chen, and Julio Pacio. "Adventures in study of the pressure losses in wire-wrapped rod bundle arrays." Nuclear Engineering and Design 370 (December 2020): 110910. http://dx.doi.org/10.1016/j.nucengdes.2020.110910.

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25

Wang, Han, Qincheng Bi, and Laurence K. H. Leung. "Heat transfer from a 2 × 2 wire-wrapped rod bundle to supercritical pressure water." International Journal of Heat and Mass Transfer 97 (June 2016): 486–501. http://dx.doi.org/10.1016/j.ijheatmasstransfer.2016.02.036.

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26

Childs, Mason, Rodolfo Vaghetto, Philip Jones, Nolan Goth, and Yassin Hassan. "Experimental determination and analysis of the transverse pressure difference in a wire-wrapped rod bundle." International Journal of Heat and Mass Transfer 170 (May 2021): 120958. http://dx.doi.org/10.1016/j.ijheatmasstransfer.2021.120958.

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27

Chai, Xiang, Xiaojing Liu, Jinbiao Xiong, and Xu Cheng. "Numerical investigation of thermal-hydraulic behaviors in a LBE-cooled 19-pin wire-wrapped rod bundle." Progress in Nuclear Energy 119 (January 2020): 103044. http://dx.doi.org/10.1016/j.pnucene.2019.103044.

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28

Chai, Xiang, Xiaojing Liu, Jinbiao Xiong, and Xu Cheng. "CFD analysis of flow blockage phenomena in a LBE-cooled 19-pin wire-wrapped rod bundle." Nuclear Engineering and Design 344 (April 2019): 107–21. http://dx.doi.org/10.1016/j.nucengdes.2019.01.019.

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29

Cesna, ? ? "Experimental Study of Heat Transfer from a Rod Bundle of Wire-Wrapped Tubes in Axial Air Flow." Heat Transfer Research 35, no. 7-8 (2004): 549–62. http://dx.doi.org/10.1615/heattransres.v35.i78.80.

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30

Su, Xing-Kang, Xian-Wen Li, Lu Zhang, Long Gu, and Da-Jun Fan. "Thermal-hydraulic study in a wire-wrapped 19-rod bundle based on an isotropic four-equation model." Annals of Nuclear Energy 178 (December 2022): 109343. http://dx.doi.org/10.1016/j.anucene.2022.109343.

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31

Wang, Shunqi, Han Wang, Jinguang Zang, Junfeng Wang, and Yanping Huang. "Numerical simulation of the thermal-hydraulic characteristics of supercritical carbon dioxide in a wire-wrapped rod bundle." Applied Thermal Engineering 198 (November 2021): 117443. http://dx.doi.org/10.1016/j.applthermaleng.2021.117443.

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32

Fernandez y Fernandez, Elói, and Pedro Carajilescov. "Static pressure and wall shear stress distributions in air flow in a seven wire-wrapped rod bundle." Journal of the Brazilian Society of Mechanical Sciences 22, no. 2 (2000): 291–302. http://dx.doi.org/10.1590/s0100-73862000000200012.

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33

Kiss, Attila, and Bence Mervay. "Numerical analysis on the thermal hydraulic effect of wrapped wire spacer in a four rod fuel bundle." Nuclear Engineering and Design 342 (February 2019): 276–307. http://dx.doi.org/10.1016/j.nucengdes.2018.11.024.

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34

Zhong, Yunke, Lian Hu, Deqi Chen, Haidong Liu, Dewen Yuan, and Wenxing Liu. "CFD simulation on the flow and heat transfer characteristics of mist flow in wire-wrapped rod bundle." Nuclear Engineering and Design 345 (April 2019): 62–73. http://dx.doi.org/10.1016/j.nucengdes.2019.02.003.

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35

Bovati, Octavio, and Yassin Hassan. "Analysis of the turbulent flow in a partially blocked wire-wrapped rod bundle using LES with wall functions." International Journal of Heat and Fluid Flow 97 (October 2022): 109041. http://dx.doi.org/10.1016/j.ijheatfluidflow.2022.109041.

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36

Liang, Yu, Dalin Zhang, Yutong Chen, Kui Zhang, Wenxi Tian, Suizheng Qiu, and Guanghui Su. "An experiment study of pressure drop and flow distribution in subchannels of a 37-pin wire-wrapped rod bundle." Applied Thermal Engineering 174 (June 2020): 115283. http://dx.doi.org/10.1016/j.applthermaleng.2020.115283.

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37

Hu, Z. X., H. B. Li, J. Q. Tao, D. Liu, and H. Y. Gu. "Experimental study on heat transfer of supercritical water flowing upward and downward in 2 × 2 rod bundle with wrapped wire." Annals of Nuclear Energy 111 (January 2018): 50–58. http://dx.doi.org/10.1016/j.anucene.2017.08.042.

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38

Chang, Seok-Kyu, Dong-Jin Euh, Hae Seob Choi, Hyungmo Kim, Sun Rock Choi, and Hyeong-Yeon Lee. "Flow Distribution and Pressure Loss in Subchannels of a Wire-Wrapped 37-pin Rod Bundle for a Sodium-Cooled Fast Reactor." Nuclear Engineering and Technology 48, no. 2 (April 2016): 376–85. http://dx.doi.org/10.1016/j.net.2015.12.013.

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39

Podila, Krishna, and Yanfei Rao. "Computational Fluid Dynamic Simulations of Heat Transfer From a 2 × 2 Wire-Wrapped Fuel Rod Bundle to Supercritical Pressure Water." Journal of Nuclear Engineering and Radiation Science 4, no. 1 (December 4, 2017). http://dx.doi.org/10.1115/1.4037747.

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Within the Generation-IV International Forum, Canadian Nuclear Laboratories (CNL) led the conceptual fuel bundle design effort for the Canadian supercritical water cooled reactor (SCWR). The proposed fuel rod assembly for the Canadian SCWR design comprised of 64-elements with spacing between elements maintained using the wire-wrap spacers. Experimental data and correlations are not available for the fuel-assembly concept of the Canadian SCWR. To analyze the thermalhydraulic performance of the new bundle design, CNL is using computational fluid dynamics (CFD) as well as the subchannel approach. Simulations of wire-wrapped bundles can benefit from the increased fidelity and resolution of a CFD approach due to its ability to resolve the boundary layer phenomena. Prior to the application, the CFD tool has been assessed against experimental heat transfer data obtained with bundle subassemblies to identify the appropriate turbulence model to use in the analyses. In the present paper, assessment of CFD predictions was made with the wire-wrapped bundle experiments performed at Xi'an Jiaotong University (XJTU) in China. A three-dimensional CFD study of the fluid flow and heat transfer at supercritical pressures for the rod-bundle geometries was performed with the key parameter being the fuel rod wall temperature. This investigation used Reynolds-averaged Navier–Stokes turbulence models with wall functions to investigate the behavior of flow through the wire-wrapped fuel rod bundles with water subjected to a supercritical pressure of 25 MPa. Along with the selection of turbulence models, CFD results were found to be dependent on the value of turbulent Prandtl number used in simulating the experimental test conditions for the wire-wrapped fuel rod configuration. It was found that the CFD simulation tends to overpredict the fuel wall temperature, and the predicted location of peak temperature differs from the measurement by up to 65 deg.

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40

Menezes, Craig, Rodolfo Vaghetto, and Yassin A. Hassan. "Experimental Investigation of the Subchannel Axial Pressure Drop and Hydraulic Characteristics of a 61-Pin Wire Wrapped Rod Bundle." Journal of Fluids Engineering 144, no. 5 (January 12, 2022). http://dx.doi.org/10.1115/1.4052745.

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Abstract Wire-wrapped hexagonal fuel bundles have been extensively investigated due to their enhanced heat transfer and flow characteristics. Experimental measurements are important to study the thermal-hydraulic behavior of such assemblies and to validate and improve the predictive capabilities of specialized correlations and computational tools. Presently, very limited experimental data is available on the local subchannel pressure drop. Experimental measurements of subchannel pressure drop were conducted in a 61-pin wire-wrapped rod bundle replica, for Reynolds numbers between 190 and 22,000. Specialized instrumented rods were utilized to measure the local pressure drop and estimate the subchannels' friction factor. Three interior subchannels, one edge subchannel, and one corner subchannel were selected to study the effects of location and flow regimes on the friction factor and hydraulic behavior. The transition boundaries from laminar to transitions regimes, and from transition to turbulent regimes were estimated for the subchannels analyzed. The results were found to be in agreement with the predictions of the upgraded Cheng and Todreas detailed correlation (UCTD). The results of the experimental campaign provided a better understanding of the hydraulic behavior of the subchannels of wire-wrapped bundles, in relation to its geometrical features.

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41

Childs, Mason, Robert Muyshondt, Rodolfo Vaghetto, Duy Thien Nguyen, and Yassin Hassan. "Experimental Study on the Effect of Localized Blockages on the Friction Factor of a 61-Pin Wire-Wrapped Bundle." Journal of Fluids Engineering 142, no. 11 (September 4, 2020). http://dx.doi.org/10.1115/1.4048140.

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Abstract The thermal-hydraulic behavior of the flow in rod bundles has motivated numerous experimental and computational investigations. Previous studies have identified potential for accumulation of debris within the small subchannels of typical wire-wrapped assemblies with subsequent total or partial blockage of subchannel coolant flow. A test campaign was conducted to study the effects of localized blockages on the bundle averaged friction factor of a tightly packed wire-wrapped rod bundle. Blockages were installed within the bundle, and fluid pressure drop was measured across one wire pitch for a Reynolds number range of 500–17,200. The Darcy–Weisbach friction factor of the perturbed rod bundle geometry was compared with that of the unblocked bundle, as well as with the predictions of a well-established friction factor correlation. Differing effects based on blockage size and location for various flow regimes were studied. A number of conclusions can be made about the effects of the blockages on the friction factor, such as an increasing effect of the blockage on friction factor with an increase in Reynolds number, a change in flow behavior in the turbulent transition flow regime near Reynolds number 3000, differences in effect on friction factor for different types of subchannel blockage, and a nonlinear trend in friction factor variation with flow area impeded for edge subchannels. To this end, all data and quantified uncertainty produced in this study are made available for comparison and validation of advanced computational tools.

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42

Kim, Hansol, Joseph Seo, and Yassin Hassan. "Prediction of flow regime boundary and pressure drop for hexagonal wire-wrapped rod bundles using artificial neural networks." Physics of Fluids, September 8, 2022. http://dx.doi.org/10.1063/5.0110385.

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This study used an artificial neural network (ANN) regression model in wire-wrapped fuel assemblies to estimate the transition-to-turbulence flow regime boundary (RebT) and friction factor. The ANN models were trained and validated using existing experimental datasets. The bundle dataset comprised several design parameters, such as the number of rods, rod diameter, wire diameter, lattice pitch, edge pitch, and wire helical pitch. The log-log scale Reynolds number and linearity characteristics of the friction coefficient were used to over-sample the friction factor in the laminar and turbulent regimes for resolving the data imbalance. Three-quarters of the entire dataset was used for training, while the remainder was used for validation. The Levenberg-Marquardt approach with the Gauss-Newton approximation for the Hessian of the training cost function was used for training the model. The number of hidden layers for RebT was selected based on the minimum validation error. The pin number effect was additionally considered for the friction factor while selecting the number of hidden layers. The ANN model predicted using the oversampled data set had a 50% reduction in root mean square error (RMSE) than the model predicted using the original data set. Compared to previous correlations, the prediction of ANN models for friction factor demonstrated significantly low errors (0.10% mean error and 7.36% RMSE of 142 bundle data).

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43

Pacio, J., S. K. Chen, Y. M. Chen, and N. E. Todreas. "Corrigendum to “Analysis of pressure losses and flow distribution in wire-wrapped hexagonal rod bundles for licensing. Part I: the Pacio-Chen-Todreas Detailed model (PCTD)” [Nucl. Eng. Des. 388 (2022) 111607]." Nuclear Engineering and Design, June 2022, 111846. http://dx.doi.org/10.1016/j.nucengdes.2022.111846.

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44

Wang, Han, Muchuan Sun, Dajun Fan, Hao Hu, Peiqi Liang, and Jianguo Yan. "Experimental Investigation of the Crossflow of Water in a 7-Pin Wire-Wrapped Rod Bundle." SSRN Electronic Journal, 2022. http://dx.doi.org/10.2139/ssrn.4086699.

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45

Menezes, Craig, Trevor Melsheimer, Dalton W. Pyle, Matthew Kinsky, and Yassin Hassan. "Flow Characteristics within an Interior Subchannel of a 61-pin Wire-Wrapped Hexagonal Rod Bundle with a Porous Blockage." Physics of Fluids, January 15, 2023. http://dx.doi.org/10.1063/5.0138487.

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Abstract:

Potential accumulation of undesirable debris in a subchannel of a Liquid Metal Fast Reactor (LMFR) hexagonal fuel bundle presents accident conditions, which are crucial to investigate. Very limited experimental research persists in literature to understand the fluid dynamics effects of partially blocked subchannels, due to the presence of porous blockages. It is imperative to comprehend flow regime-dependent fluid response in the vicinity of porous blockages, to predict and counter abnormal conditions in an LMFR rod assembly. The presented experimental research investigates flow-field characteristics in a 61-pin wire-wrapped rod assembly with a three-dimensional (3D) printed porous blockage medium in an interior subchannel, at Reynolds numbers (Re) of 350, 5,000, and 14,000. Time-resolved velocimetry measurements were acquired yielding first- and second-order Reynolds decomposition flow statistics - revealing important fluid responses upstream and downstream of the porous blockage. Profiles of velocities, velocity fluctuations, Reynolds stresses, and vorticities uncovered the downstream blockage perturbation effects. Spatial cross-correlations of the velocity fluctuations displayed eddie structure elongations and quantified eddie integral scale lengths. A time-frequency analysis of the velocity fluctuations further detailed the mechanisms of flow instabilities via power spectral analysis. Application of a one-dimensional continuous wavelet transform revealed complex Re-dependent flow and characterized the temporal turbulence occurrences - caused by the trailing edge effects of the porous blockage. This research provides unique and novel experimental analyses on flow regime-dependent fluid physics due to a porous blockage medium and provides data sets vital for computational model benchmarking and development, towards the enhancement of LMFR rod bundle designs.

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46

Dong, Kejian, Shakeel Ahmad, Shahid Ali Khan, Peng Ding, Wenhuai Li, and Jiyun Zhao. "Thermal‐hydraulic analysis of wire‐wrapped rod bundle in lead‐based fast reactor with non‐uniform heat flux." International Journal of Energy Research, June 28, 2022. http://dx.doi.org/10.1002/er.8316.

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47

Quan, Zhengting, Adam Dix, Ran Kong, Seungjin Kim, Mamoru Ishii, and Mitchell T. Farmer. "Pressure Drop in Seven-Pin Wire-Wrapped Rod Bundle for the Sodium Cartridge Loop in Versatile Test Reactor." Nuclear Science and Engineering, July 5, 2022, 1–17. http://dx.doi.org/10.1080/00295639.2022.2082232.

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