Core thermal-hydraulic design

Eiji Takada, Shigeaki Nakagawa, Nozomu Fujimoto, Daisuke Tochio

Research output: Contribution to journalArticle

13 Citations (Scopus)

Abstract

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.

Original languageEnglish
Pages (from-to)37-43
Number of pages7
JournalNuclear Engineering and Design
Volume233
Issue number1-3
DOIs
Publication statusPublished - Oct 1 2004
Externally publishedYes

Fingerprint

hydraulics
Hydraulics
temperature
Temperature
coolants
Coolants
high temperature tests
occurrences
test facilities
Test facilities
Hot Temperature
heat transfer coefficients
Heat transfer coefficients
heat transfer
safety
flow distribution
cycles
evaluation
coefficients

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

Core thermal-hydraulic design. / Takada, Eiji; Nakagawa, Shigeaki; Fujimoto, Nozomu; Tochio, Daisuke.

In: Nuclear Engineering and Design, Vol. 233, No. 1-3, 01.10.2004, p. 37-43.

Research output: Contribution to journalArticle

Takada, Eiji ; Nakagawa, Shigeaki ; Fujimoto, Nozomu ; Tochio, Daisuke. / Core thermal-hydraulic design. In: Nuclear Engineering and Design. 2004 ; Vol. 233, No. 1-3. pp. 37-43.
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