Evaluation of core thermal and hydraulic characteristics of HTTR

So Maruyama, Nozomu Fujimoto, Yukio Sudo, Tomoyuki Murakami, Sadao Fujii

研究成果: Contribution to journalArticle査読

17 被引用数 (Scopus)

抄録

Japan Atomic Energy Research Institute has started the development of the high temperature engineering test reactor (HTTR), a graphite-moderated, helium gas-cooled reactor with 30 MW thermal power and maximum outlet coolant temperature of 950 °C. This paper describes the core thermal and hydraulic (T/H) design procedure, including the validation of the computer code system, design criteria pertaining to the fuel design limit and the evaluated core T/H charateristics. The core T/H design of the HTTR has been carried out considering the specific characteristics of the core structure and the fuel based on R&D results. The coolant flow rate and temperature distribution are evaluated by the flow network analysis code flownet. The fuel temperature distribution is evaluated by the fuel temperature analysis code temdim with multi-cylindrical model using hot spot factors. Fuel design limit for anticipated operational occurrences and fuel temperature limit for normal operation are specified at 1600°C and 1495°C, respectively based on experimental results. Several design considerations are also adopted to realize a high reactor outlet coolant temperature of 950°C. As a result of core T/H design, the effective core flow rate and maximum fuel temperature during the high temperature test operation are 88% and 1492°C, respectively.

本文言語英語
ページ(範囲)183-196
ページ数14
ジャーナルNuclear Engineering and Design
152
1-3
DOI
出版ステータス出版済み - 11 3 1994
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

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