Core thermal-hydraulic design

Eiji Takada, Shigeaki Nakagawa, Nozomu Fujimoto, Daisuke Tochio

研究成果: ジャーナルへの寄稿学術誌査読

17 被引用数 (Scopus)

抄録

The core thermal-hydraulic design for the HTTR is carried out to evaluate the maximum fuel temperature at normal operation and anticipated operation occurrences. To evaluate coolant flow distribution and maximum fuel temperature, we use the experimental results such as heat transfer coefficient, pressure loss coefficient obtained by mock-up test facilities. Furthermore, we evaluated hot spot factors of fuel temperatures conservatively. As the results of the core thermal-hydraulic design, an effective coolant flow through the core of 88% of the total flow, is achieved at minimum. The maximum fuel temperature appears during the high-temperature test operation, and reaches 1492 °C for the maximum through the burn-up cycle, which satisfies the design limit of 1495 °C at normal operation. It is also confirmed that the maximum fuel temperature at any anticipated operation occurrences does not exceed the fuel design limit of 1600 °C in the safety analysis. On the other hand, result of re-evaluation of analysis condition and hot spot factors based on operation data of the HTTR, the maximum fuel temperature for 160 effective full power operation days is estimated to be 1463 °C. It is confirmed that the core thermal-hydraulic design gives conservative results.

本文言語英語
ページ(範囲)37-43
ページ数7
ジャーナルNuclear Engineering and Design
233
1-3
DOI
出版ステータス出版済み - 10月 1 2004
外部発表はい

!!!All Science Journal Classification (ASJC) codes

  • 核物理学および高エネルギー物理学
  • 原子力エネルギーおよび原子力工学
  • 材料科学(全般)
  • 安全性、リスク、信頼性、品質管理
  • 廃棄物管理と処理
  • 機械工学

フィンガープリント

「Core thermal-hydraulic design」の研究トピックを掘り下げます。これらがまとまってユニークなフィンガープリントを構成します。

引用スタイル