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Title: Critical Heat Flux Experiments on the Reactor Vessel Wall Using 2-D Slice Test Section

Abstract

The critical heat flux (CHF) on the reactor vessel outer wall was measured using the two-dimensional slice test section. The radius and the channel area of the test section were 2.5 m and 10 cm x 15 cm, respectively. The flow channel area and the heater width were smaller than those of the ULPU experiments, but the radius was greater than that of the ULPU. The CHF data under the inlet subcooling of 2 to 25 deg. C and the mass flux 0 to 300 kg/m{sup 2}.s had been acquired. The measured CHF value was generally slightly lower than that of the ULPU. The difference possibly comes from the difference of the test section material and the thickness. However, the general trend of CHF according to the mass flux was similar with that of the ULPU. The experimental CHF data were compared with the predicted values by SULTAN correlation. The SULTAN correlation predicted well this study's data only for the mass flux higher than 200 kg/m{sup 2}.s, and for the exit quality lower than 0.05. The local condition-based correlation was developed, and it showed good prediction capability for broad quality (-0.01 to 0.5) and mass flux (<300 kg/m{sup 2}.s) conditionsmore » with a root-mean-square error of 2.4%. There were increases in the CHF with trisodium phosphate-added water.« less

Authors:
 [1];  [1];  [2]
  1. Korea Advanced Institute of Science and Technology (Korea, Republic of)
  2. Korea Atomic Energy Research Institute (Korea, Republic of)
Publication Date:
OSTI Identifier:
20840318
Resource Type:
Journal Article
Resource Relation:
Journal Name: Nuclear Technology; Journal Volume: 152; Journal Issue: 2; Other Information: Copyright (c) 2006 American Nuclear Society (ANS), United States, All rights reserved. http://epubs.ans.org/; Country of input: International Atomic Energy Agency (IAEA)
Country of Publication:
United States
Language:
English
Subject:
22 GENERAL STUDIES OF NUCLEAR REACTORS; CRITICAL HEAT FLUX; HEATERS; MASS; PHOSPHATES; REACTOR VESSELS; SUBCOOLING; THICKNESS; TWO-DIMENSIONAL CALCULATIONS; WALLS; WATER; WIDTH

Citation Formats

Jeong, Yong Hoon, Chang, Soon Heung, and Baek, Won-Pil. Critical Heat Flux Experiments on the Reactor Vessel Wall Using 2-D Slice Test Section. United States: N. p., 2005. Web.
Jeong, Yong Hoon, Chang, Soon Heung, & Baek, Won-Pil. Critical Heat Flux Experiments on the Reactor Vessel Wall Using 2-D Slice Test Section. United States.
Jeong, Yong Hoon, Chang, Soon Heung, and Baek, Won-Pil. Tue . "Critical Heat Flux Experiments on the Reactor Vessel Wall Using 2-D Slice Test Section". United States. doi:.
@article{osti_20840318,
title = {Critical Heat Flux Experiments on the Reactor Vessel Wall Using 2-D Slice Test Section},
author = {Jeong, Yong Hoon and Chang, Soon Heung and Baek, Won-Pil},
abstractNote = {The critical heat flux (CHF) on the reactor vessel outer wall was measured using the two-dimensional slice test section. The radius and the channel area of the test section were 2.5 m and 10 cm x 15 cm, respectively. The flow channel area and the heater width were smaller than those of the ULPU experiments, but the radius was greater than that of the ULPU. The CHF data under the inlet subcooling of 2 to 25 deg. C and the mass flux 0 to 300 kg/m{sup 2}.s had been acquired. The measured CHF value was generally slightly lower than that of the ULPU. The difference possibly comes from the difference of the test section material and the thickness. However, the general trend of CHF according to the mass flux was similar with that of the ULPU. The experimental CHF data were compared with the predicted values by SULTAN correlation. The SULTAN correlation predicted well this study's data only for the mass flux higher than 200 kg/m{sup 2}.s, and for the exit quality lower than 0.05. The local condition-based correlation was developed, and it showed good prediction capability for broad quality (-0.01 to 0.5) and mass flux (<300 kg/m{sup 2}.s) conditions with a root-mean-square error of 2.4%. There were increases in the CHF with trisodium phosphate-added water.},
doi = {},
journal = {Nuclear Technology},
number = 2,
volume = 152,
place = {United States},
year = {Tue Nov 15 00:00:00 EST 2005},
month = {Tue Nov 15 00:00:00 EST 2005}
}