Countercurrent flow limitation in thin rectangular channels between advanced test reactor fuel elements
Conference
·
· Transactions of the American Nuclear Society; (United States)
OSTI ID:6659372
- Idaho National Engineering Lab., Idaho Falls (United States)
An analytical and experimental investigation of countercurrent flow limitation (CCFL) in thin, rectangular channels, similar to those between adjacent fuel elements in the Advanced Test Reactor (ATR), was conducted at the Idaho National Entgineering Laboratory. The ATR is a 250-MW(thermal) materials irradiation facility operated by EG G Idaho, Inc. for the US Department of Energy. During a postulated large-break loss-of-coolant accident (LOCA), the ATRs emergency core cooling system will inject water directly into the reactor vessel both above and below the core. However, reflood via the upper injection system may be limited by upward flowing steam. Flooding experiments were carried out with air and water to evaluate the flooding limits in a 0.0011- (or 0.0022)- [times] 0.064- [times] 1.55-m test section. The test section represented the flow channel between the side plates of adjacent ATR fuel elements through which coolant could enter the core during a large-break LOCA. The objectives of the present study were to (a) develop a drift-flux CCFL model from the experimental data and (b) assess the ability of the RELAP5 thermal-hydraulic code to simulate CCFL.
- OSTI ID:
- 6659372
- Report Number(s):
- CONF-921102--
- Conference Information:
- Journal Name: Transactions of the American Nuclear Society; (United States) Journal Volume: 66
- Country of Publication:
- United States
- Language:
- English
Similar Records
Advanced test reactor countercurrent flow experiment
Mitigation of countercurrent flow limiting at the core outlet during a LOCA
RELAP5 simulation of the ROSA-IV/LSTF SB-CL-18 test using countercurrent flow limitation modeling
Conference
·
Fri Jun 01 00:00:00 EDT 1990
· Transactions of the American Nuclear Society; (USA)
·
OSTI ID:6371584
Mitigation of countercurrent flow limiting at the core outlet during a LOCA
Conference
·
Wed Dec 31 23:00:00 EST 1986
· Trans. Am. Nucl. Soc.; (United States)
·
OSTI ID:6942589
RELAP5 simulation of the ROSA-IV/LSTF SB-CL-18 test using countercurrent flow limitation modeling
Conference
·
Thu Dec 31 23:00:00 EST 1992
· Transactions of the American Nuclear Society; (United States)
·
OSTI ID:7129137
Related Subjects
21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS
22 GENERAL STUDIES OF NUCLEAR REACTORS
220600 -- Nuclear Reactor Technology-- Research
Test & Experimental Reactors
220900* -- Nuclear Reactor Technology-- Reactor Safety
ACCIDENTS
ATR REACTOR
COMPUTER CODES
ECCS
ENGINEERED SAFETY SYSTEMS
ENRICHED URANIUM REACTORS
FLOW BLOCKAGE
FLOW MODELS
FLUID FLOW
FLUID MECHANICS
FUEL ASSEMBLIES
FUEL CHANNELS
FUEL ELEMENTS
HYDRAULICS
IRRADIATION REACTORS
LOSS OF COOLANT
MATERIALS TESTING REACTORS
MATHEMATICAL MODELS
MECHANICS
R CODES
REACTOR ACCIDENTS
REACTOR CHANNELS
REACTOR COMPONENTS
REACTOR PROTECTION SYSTEMS
REACTORS
RESEARCH AND TEST REACTORS
STEAM
TANK TYPE REACTORS
TEST REACTORS
TESTING
THERMAL REACTORS
WATER COOLED REACTORS
WATER MODERATED REACTORS
22 GENERAL STUDIES OF NUCLEAR REACTORS
220600 -- Nuclear Reactor Technology-- Research
Test & Experimental Reactors
220900* -- Nuclear Reactor Technology-- Reactor Safety
ACCIDENTS
ATR REACTOR
COMPUTER CODES
ECCS
ENGINEERED SAFETY SYSTEMS
ENRICHED URANIUM REACTORS
FLOW BLOCKAGE
FLOW MODELS
FLUID FLOW
FLUID MECHANICS
FUEL ASSEMBLIES
FUEL CHANNELS
FUEL ELEMENTS
HYDRAULICS
IRRADIATION REACTORS
LOSS OF COOLANT
MATERIALS TESTING REACTORS
MATHEMATICAL MODELS
MECHANICS
R CODES
REACTOR ACCIDENTS
REACTOR CHANNELS
REACTOR COMPONENTS
REACTOR PROTECTION SYSTEMS
REACTORS
RESEARCH AND TEST REACTORS
STEAM
TANK TYPE REACTORS
TEST REACTORS
TESTING
THERMAL REACTORS
WATER COOLED REACTORS
WATER MODERATED REACTORS