Boiled-up level and boiling two-phase flow dynamics in 5 × 5 heated rod bundle during boil-off process under atmospheric pressure conditions

Takahiro Arai, Masahiro Furuya, Taizo Kanai, Kenetsu Shirakawa, Yoshihisa Nishi

研究成果: Conference contribution

2 被引用数 (Scopus)

抄録

In the case of an accident and when a water level of a reactor core falls below the top level of active fuel, the cooling limit height becomes a key factor for determining the accident mitigation procedure. To predict the cooling limit height, it is important to clarify a two-phase mixture level in a rod bundle during the boil-off process. The two-phase mixture level depends on the collapsed level and void fraction distribution. During the boil-off process, a boiling two-phase flow in the rod bundle exhibits multidimensional and complex flow structures. The paper addresses a three-dimensional void fraction distribution and a two-phase mixture level in 5 × 5 heated rod bundles during the boil-off process, under atmospheric pressure conditions. The heated rod length is 3.7 m, which is the same as the fuel rod in boiling water-reactor (BWR). The 5x5 rod bundles have an axially and radially uniform power profile, and eight pairs of sheath thermocouples are embedded in the heated rod to monitor their axial surface temperature profiles. The diameter of the heated rod is 10 mm, and the rod pitch is 13 mm. The void fraction distributions were acquired with eight pairs of subchannel void sensors (SCVS) as time series data. The two-phase mixture level was evaluated by side-viewing images acquired with two high-speed digital video cameras. The experimental result exhibits a relationship of the boiling two-phase flow dynamics to the two-phase mixture level, and the void fraction during the boil-off process.

本文言語English
ホスト出版物のタイトルInternational Topical Meeting on Nuclear Reactor Thermal Hydraulics 2015, NURETH 2015
出版社American Nuclear Society
ページ7312-7322
ページ数11
ISBN(電子版)9781510811843
出版ステータスPublished - 2015
外部発表はい
イベント16th International Topical Meeting on Nuclear Reactor Thermal Hydraulics, NURETH 2015 - Chicago, United States
継続期間: 2015 8 302015 9 4

出版物シリーズ

名前International Topical Meeting on Nuclear Reactor Thermal Hydraulics 2015, NURETH 2015
9

Other

Other16th International Topical Meeting on Nuclear Reactor Thermal Hydraulics, NURETH 2015
CountryUnited States
CityChicago
Period15/8/3015/9/4

ASJC Scopus subject areas

  • Instrumentation
  • Nuclear Energy and Engineering

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