TY - JOUR
T1 - Study of critical heat flux in natural convection-cooled TRIGA reactors with single annulus and rod bundle geometries
AU - Yang, Jun
AU - Greenwood, Michael Scott
AU - De Angelis, Matthew
AU - Avery, Michael
AU - Anderson, Mark
AU - Corradini, Michael
AU - Matos, James
AU - Dunn, Floyd
AU - Feldman, Earl
PY - 2015/6/1
Y1 - 2015/6/1
N2 - A critical heat flux (CHF) experimental study at low pressure and natural convection condition has been conducted. The test apparatus is a natural circulation loop with an upward flow channel, simulating TRIGA (Training, Research, Isotopes, General Atomics) reactors. CHF is studied in three types of geometries: a single-rod annulus, a three-rod bundle in a trefoil tube, and a four-rod bundle in a square tube. The full-scale fuel pin heater rod is electrically heated with a prototypic axial power profile, equipped with thermocouples for CHF detection. Experiments are carried out at the following conditions: inlet subcooling from 10 to 70 K, pressure from 110 to 290 kPa, and mass flux from 0 to 400 kg/m2.s. It is observed that CHF increases as the pressure or mass flux increases but does not significantly depend on the inlet subcooling within the testing range. The current CHF data are compared with a few selected CHF correlations whose application ranges are close to the testing conditions. The relevance of the CHF to the testing parameters is investigated. A modified CHF correlation compatible with TRIGA reactor conditions is proposed based on a previous correlation and current experimental data.
AB - A critical heat flux (CHF) experimental study at low pressure and natural convection condition has been conducted. The test apparatus is a natural circulation loop with an upward flow channel, simulating TRIGA (Training, Research, Isotopes, General Atomics) reactors. CHF is studied in three types of geometries: a single-rod annulus, a three-rod bundle in a trefoil tube, and a four-rod bundle in a square tube. The full-scale fuel pin heater rod is electrically heated with a prototypic axial power profile, equipped with thermocouples for CHF detection. Experiments are carried out at the following conditions: inlet subcooling from 10 to 70 K, pressure from 110 to 290 kPa, and mass flux from 0 to 400 kg/m2.s. It is observed that CHF increases as the pressure or mass flux increases but does not significantly depend on the inlet subcooling within the testing range. The current CHF data are compared with a few selected CHF correlations whose application ranges are close to the testing conditions. The relevance of the CHF to the testing parameters is investigated. A modified CHF correlation compatible with TRIGA reactor conditions is proposed based on a previous correlation and current experimental data.
UR - http://www.scopus.com/inward/record.url?scp=84931837279&partnerID=8YFLogxK
U2 - 10.13182/NSE14-45
DO - 10.13182/NSE14-45
M3 - Article
AN - SCOPUS:84931837279
SN - 0029-5639
VL - 180
SP - 141
EP - 153
JO - Nuclear Science and Engineering
JF - Nuclear Science and Engineering
IS - 2
ER -