Published June 1996 | Version v1
Journal article

Comments on the letter to the editor and some notes

  • 1. Tsukuba Univ., Ibaraki (Japan). Inst. of Eng. and Mech.
  • 2. Ship Research Inst., Tokai, Ibaraki (Japan)

Description

We have conducted our experimental research on the critical heat flux (CHF) of subcooled flow boiling with water for many years in order to make clear the coolability limit of high heat flux components of fusion reactors. The research concerned the CHF of smooth and twisted tape tubes both under circumferentially uniform and circumferentially non-uniform heat flux conditions. All experiments were conducted under axially uniform heat flux conditions. We proposed empirical correlations for smooth tube (modified Tong correlation) and for twisted tape tube (Nariai-Inasake correlation), and then conducted the evaluation of various correlations proposed to date by many researchers, including ourselves. Our evaluation results, however sometimes differed from other researchers' evaluations; we proposed that the reason for the differences might be the evaluation methods used. We called the two methods used direct substitution method (DSM) and heat balance method (HBM), and adopted HBM for the evaluation of correlations (Inasaka and Nariai, 1996). This is a comment on Todreas, N.E. and Hejzlar, P. ibid vol. 163(1-2) p. 25 (1996). (orig.)

Additional details

Publishing Information

Journal Title
Nuclear Engineering and Design
Journal Volume
163
Journal Issue
1-2
Journal Page Range
p. 281-282.
ISSN
0029-5493
CODEN
NEDEAU

INIS

Country of Publication
Netherlands
Country of Input or Organization
Switzerland
INIS RN
27068469
Subject category
S22: GENERAL STUDIES OF NUCLEAR REACTORS;
Descriptors DEI
COMPUTERIZED SIMULATION; CORRELATIONS; CRITICAL HEAT FLUX; REACTOR COOLING SYSTEMS; SUBCOOLED BOILING; TUBES
Descriptors DEC
BOILING; COOLING SYSTEMS; HEAT FLUX; PHASE TRANSFORMATIONS; REACTOR COMPONENTS; SIMULATION