Results 11 to 20 of about 361,273 (254)
Nuclear Engineering and TechnologyStudy of flow behavior within rod bundles has been an active topic. Surface modification technologies are important parts of the design of the fourth generation reactor, which can increase the strength of the secondary flow within the rod bundle lattices.
Yi Liao +6 more
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Experimental studies of a single flexibly-mounted rod in a triangular rod bundle in cross-flow [PDF]
MATEC Web of Conferences, 2018 Experiments on flow-induced vibrations using a closely-packed triangular rod array with a pitch-todiameter ratio of 1.1 in water cross-flow was carried out at Paul Scherrer Institute. The bundle consists of 21 row of five rods in each one.
Upnere Sabine +3 more
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Yuanzineng kexue jishu, 2022 There is mixing and second flow in sub-channels of the rod bundle. The flow and heat transfer characteristics in rod bundle are more complex than those in rectangular channel and round pipes.
BI Jingliang;XIE Feng;HUANG Yanping;XU Jianjun;YUAN Dewen;ZAN Yuanfeng
doaj
CFD Analysis of Scaled Wire-wrapped rod Bundle Cooled by LBE [PDF]
E3S Web of Conferences, 2019 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
LYU Kefeng +3 more
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Design concept and testing of an in-bundle gamma densitometer for subchannel void fraction measurements in the THTF electrically heated rod bundle. [PWR] [PDF]
, 1982 A design concept is presented for an in-bundle gamma densitometer system for measurement of subchannel average fluid density and void fraction in rod or tube bundles.
Felde, D. K.
core +1 more source
Mechanism Model of Critical Heat Flux in Bundle Based on High-precision Subchannel Code
Yuanzineng kexue jishu, 2023 Critical heat flux (CHF) is a crucial safety limit for PWR core design. Accurately predicting CHF in the core rod bundle channel can enhance the economy and safety of the reactor. Currently, the prediction of rod bundle CHF in engineering is mainly based
GUO Junliang;JIANG Li;GAO Kui;KONG Huanjun;GUI Miao;SHAN Jianqiang
doaj
Mechanical Engineering Journal, 2015 A subchannel void sensor (SCVS) was developed to measure the cross-sectional distribution of void fraction in a 5×5 heated rod bundle with o.d. 10 mm and heated length 2000 mm, and applied to a boiling two-phase flow experiment under the atmospheric ...
Takahiro ARAI +4 more
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Frontiers in Energy Research, 2022 Helium flow in the rod bundle channel with an open lattice structure is an important phenomenon for the advanced gas-cooled nuclear core design. In this study, thermal analysis with helium flow in various channel designs is conducted based on CFD methods
Zhipeng Wang +11 more
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Assessment of a Theoretical Model for Predicting Forced Convective Critical Heat Flux in Rod Bundles
Frontiers in Energy Research, 2019 This study proposes a theoretical model for predicting forced convective critical heat flux (CHF) in rod bundles based on the bubble crowding phenomenon. The theoretical model applied to the rod bundle is based on Weisman-Pei's basic tube model. In order
Yang Liu +5 more
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Hierarchical entropy analysis for flow pattern of steam-liquid two phase flow in a rod bundle
Frontiers in Energy Research, 2023 Flow pattern is very key parameter in the study of flow characteristics and heat transfer characteristics of two-phase flow. In this Paper, two-phase boiling experiment in a 3×3 rod bundle was conducted within the following parameter range: 0≤G≤600kg/(m2⋅
Chao Zhang +11 more
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