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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}
}
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  • The government of Mexico has expressed interest in utilizing the Laguna Verde boiling water reactor (BWR) nuclear power plant for the disposition of reprocessed spent uranium oxide (UOX) fuel in the form of reactor-grade mixed-oxide (MOX) fuel. MOX fuel would replace spent UOX fuel as a fraction in the core from 18 to 30% depending on the fuel loading cycle. MOX fuel is expected to increase the neutron fluence, flux, fuel centerline temperature, reactor core pressure, and yield higher energy neutrons.There is concern that a core with a fraction of MOX fuel (i.e., increased {sup 239}Pu wt%) would increase themore » radiation damage displacements per atom per second (dpa-s{sup -1}) in steel within the core shroud and vessel wall as compared to only conventional, enriched UOX fuel in the core. The evaluation of radiation damage within the core shroud and vessel wall is a concern because of the potentially adverse affect to personnel and public safety, environment, and operating life of the reactor.The primary uniqueness of this paper is the computation of radiation damage (dpa-s{sup -1}) using NJOY99-processed cross sections for steel within the core shroud and vessel wall. Specifically, the unique radiation damage results are several orders of magnitude greater than results of previous works. In addition, the conclusion of this paper was that the addition of the maximum fraction of one-third MOX fuel to the LV1 BWR core did significantly increase the radiation damage in steel within the core shroud and vessel wall such that without mitigation of radiation damage by periodic thermal annealing or reduction in operating parameters such as neutron fluence, core temperature, and pressure, it posed a potentially adverse affect to the personnel and public safety, environment, and operating life of the reactor.« less