Abstract
It is investigated thermal and fluid mixing phenomena in situations relevant to pressurized thermal shock in the downcomer of a PWR during a cooldown transient caused by direct vessel injection(DVI). The test model was designed to simulate the typical values of prototype Combustion Engineering System 80{sup +} by approximately factors of five. The test facility is consisted with water storage tanks, heating systems, pumps, and flowmeters needed to establish the required initial conditions and to run tests of sufficient duration to achieve steady fluid temperature through out the test vessel. Test measurements include flow rates, temperatures and pressures at the boundaries of the 1/5-scale model facility to characterize the test conditions, plus over 10 local measurements of temperature within the downcomer. Test data are recorded from all instruments simultaneously on a computer-based data acquisition system for the duration of each test. The tests were carried out to simulate two postulated accident conditions. The one was small-break loss-of-coolant accident(SBLOCA), and the other was steam line break(SLB). The results were compared with predicted values calculated with REMIX-code and COMMIX-code. The compared results could corroborate the trend of thermal and fluid mixing during cooldown transient with the result of visualization tests. Analysis of
More>>
Cha, Jong Hee;
Chung, Moon Ki;
Won, Soon Yeon;
Jun, Hyung Gil
[1]
- Korea Atomic Energy Research Inst., Daeduk (Korea, Republic of)
Citation Formats
Cha, Jong Hee, Chung, Moon Ki, Won, Soon Yeon, and Jun, Hyung Gil.
The thermal-hydraulic characteristics of the nuclear reactor core during accident -an investigation on thermal mixing with direct vessel injection(II)-.
Korea, Republic of: N. p.,
1991.
Web.
Cha, Jong Hee, Chung, Moon Ki, Won, Soon Yeon, & Jun, Hyung Gil.
The thermal-hydraulic characteristics of the nuclear reactor core during accident -an investigation on thermal mixing with direct vessel injection(II)-.
Korea, Republic of.
Cha, Jong Hee, Chung, Moon Ki, Won, Soon Yeon, and Jun, Hyung Gil.
1991.
"The thermal-hydraulic characteristics of the nuclear reactor core during accident -an investigation on thermal mixing with direct vessel injection(II)-."
Korea, Republic of.
@misc{etde_10113180,
title = {The thermal-hydraulic characteristics of the nuclear reactor core during accident -an investigation on thermal mixing with direct vessel injection(II)-}
author = {Cha, Jong Hee, Chung, Moon Ki, Won, Soon Yeon, and Jun, Hyung Gil}
abstractNote = {It is investigated thermal and fluid mixing phenomena in situations relevant to pressurized thermal shock in the downcomer of a PWR during a cooldown transient caused by direct vessel injection(DVI). The test model was designed to simulate the typical values of prototype Combustion Engineering System 80{sup +} by approximately factors of five. The test facility is consisted with water storage tanks, heating systems, pumps, and flowmeters needed to establish the required initial conditions and to run tests of sufficient duration to achieve steady fluid temperature through out the test vessel. Test measurements include flow rates, temperatures and pressures at the boundaries of the 1/5-scale model facility to characterize the test conditions, plus over 10 local measurements of temperature within the downcomer. Test data are recorded from all instruments simultaneously on a computer-based data acquisition system for the duration of each test. The tests were carried out to simulate two postulated accident conditions. The one was small-break loss-of-coolant accident(SBLOCA), and the other was steam line break(SLB). The results were compared with predicted values calculated with REMIX-code and COMMIX-code. The compared results could corroborate the trend of thermal and fluid mixing during cooldown transient with the result of visualization tests. Analysis of test results and predicted values may lead to the following conclusion. The fluid and thermal mixing pattern may be devided into three different regions, which are the flume flow region, the mixing developing region and fully mixed region. (Author).}
place = {Korea, Republic of}
year = {1991}
month = {Jan}
}
title = {The thermal-hydraulic characteristics of the nuclear reactor core during accident -an investigation on thermal mixing with direct vessel injection(II)-}
author = {Cha, Jong Hee, Chung, Moon Ki, Won, Soon Yeon, and Jun, Hyung Gil}
abstractNote = {It is investigated thermal and fluid mixing phenomena in situations relevant to pressurized thermal shock in the downcomer of a PWR during a cooldown transient caused by direct vessel injection(DVI). The test model was designed to simulate the typical values of prototype Combustion Engineering System 80{sup +} by approximately factors of five. The test facility is consisted with water storage tanks, heating systems, pumps, and flowmeters needed to establish the required initial conditions and to run tests of sufficient duration to achieve steady fluid temperature through out the test vessel. Test measurements include flow rates, temperatures and pressures at the boundaries of the 1/5-scale model facility to characterize the test conditions, plus over 10 local measurements of temperature within the downcomer. Test data are recorded from all instruments simultaneously on a computer-based data acquisition system for the duration of each test. The tests were carried out to simulate two postulated accident conditions. The one was small-break loss-of-coolant accident(SBLOCA), and the other was steam line break(SLB). The results were compared with predicted values calculated with REMIX-code and COMMIX-code. The compared results could corroborate the trend of thermal and fluid mixing during cooldown transient with the result of visualization tests. Analysis of test results and predicted values may lead to the following conclusion. The fluid and thermal mixing pattern may be devided into three different regions, which are the flume flow region, the mixing developing region and fully mixed region. (Author).}
place = {Korea, Republic of}
year = {1991}
month = {Jan}
}