Evaluation of core thermal and hydraulic characteristics of HTTR

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

Research output: Contribution to journalArticle

14 Citations (Scopus)

Abstract

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.

Original languageEnglish
Pages (from-to)183-196
Number of pages14
JournalNuclear Engineering and Design
Volume152
Issue number1-3
DOIs
Publication statusPublished - Nov 3 1994
Externally publishedYes

Fingerprint

engineering test reactors
High temperature engineering
hydraulics
Hydraulics
engineering
evaluation
coolants
Coolants
temperature
outlets
Temperature
Temperature distribution
temperature distribution
flow velocity
high temperature tests
Flow rate
gas cooled reactors
core flow
Gas cooled reactors
network analysis

All Science Journal Classification (ASJC) codes

  • Nuclear and High Energy Physics
  • Nuclear Energy and Engineering
  • Materials Science(all)
  • Safety, Risk, Reliability and Quality
  • Waste Management and Disposal
  • Mechanical Engineering

Cite this

Evaluation of core thermal and hydraulic characteristics of HTTR. / Maruyama, So; Fujimoto, Nozomu; Sudo, Yukio; Murakami, Tomoyuki; Fujii, Sadao.

In: Nuclear Engineering and Design, Vol. 152, No. 1-3, 03.11.1994, p. 183-196.

Research output: Contribution to journalArticle

Maruyama, So ; Fujimoto, Nozomu ; Sudo, Yukio ; Murakami, Tomoyuki ; Fujii, Sadao. / Evaluation of core thermal and hydraulic characteristics of HTTR. In: Nuclear Engineering and Design. 1994 ; Vol. 152, No. 1-3. pp. 183-196.
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