You need JavaScript to view this

The thermal-hydraulic characteristics of the nuclear reactor core during accident -an investigation on thermal mixing with direct vessel injection(II)-

Technical Report:

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>>
Authors:
Cha, Jong Hee; Chung, Moon Ki; Won, Soon Yeon; Jun, Hyung Gil [1] 
  1. Korea Atomic Energy Research Inst., Daeduk (Korea, Republic of)
Publication Date:
Jan 01, 1991
Product Type:
Technical Report
Report Number:
KAERI/RR-960/90
Reference Number:
SCA: 220200; 210200; PA: AIX-23:014924; SN: 92000646922
Resource Relation:
Other Information: PBD: Jan 1991
Subject:
22 GENERAL STUDIES OF NUCLEAR REACTORS; 21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS; REACTOR CORES; HYDRODYNAMICS; THERMODYNAMIC PROPERTIES; ACCIDENTS; INJECTION; REACTOR VESSELS; TEST FACILITIES; THERMAL SHOCK; 220200; 210200; COMPONENTS AND ACCESSORIES; POWER REACTORS, NONBREEDING, LIGHT-WATER MODERATED, NONBOILING WATER COOLED
OSTI ID:
10113180
Research Organizations:
Korea Atomic Energy Research Inst., Daeduk (Korea, Republic of)
Country of Origin:
Korea, Republic of
Language:
Korean
Other Identifying Numbers:
Other: ON: DE92615089; TRN: KR9100114014924
Availability:
OSTI; NTIS (US Sales Only); INIS
Submitting Site:
INIS
Size:
134 p.
Announcement Date:
Jun 30, 2005

Technical Report:

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}
}