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Multiphase Science and Technology

年間 4 号発行

ISSN 印刷: 0276-1459

ISSN オンライン: 1943-6181

SJR: 0.144 SNIP: 0.256 CiteScore™:: 1.1 H-Index: 24

Indexed in

MODEL OF THE COOLING OF A NUCLEAR REACTOR FUEL ROD

巻 25, 発行 2-4, 2013, pp. 237-248
DOI: 10.1615/MultScienTechn.v25.i2-4.90
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要約

This paper presents an experimental setup showing some typical phenomena associated with the cooling of a fuel rod inside a nuclear reactor. The fuel rod model allows the general public to observe phenomena related to fluid mechanics and heat transfer. Safety and operational reasons usually prevent these phenomena from being observed in real-life equipment. The most interesting and important phenomena are boiling crisis and transient heat transfer in two-phase flow.

によって引用された
  1. Polansky Jiri, Wang Mi, Vertical annular flow pattern characterisation using proper orthogonal decomposition of Electrical Impedance Tomography, Flow Measurement and Instrumentation, 62, 2018. Crossref

  2. Silvi Liril D., Chandraker Dinesh K., Ghosh Sumana, Das Arup K, Understanding dry-out mechanism in rod bundles of boiling water reactor, International Journal of Heat and Mass Transfer, 177, 2021. Crossref

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