RELAP5/MOD2 simulation of ORNL reflood tests
Conference
·
· Trans. Am. Nucl. Soc.; (United States)
OSTI ID:7118620
Following the blowdown phase of a postulated loss-of-coolant accident in a pressurized water reactor (PWR), the reactor core is uncovered and the fuel rods experience rapid temperature excursions because of decay power production and low heat transfer. In order to prevent fuel from overheating, emergency core cooling is activated and the accident enters the reflood phase. Several experimental and analytical programs have been performed to investigate the reflooding phenomena. The purpose of this study is to simulate several reflooding tests performed at Oak Ridge National Laboratory in the Thermal-Hydraulic Test Facility (THTF) using the latest version of RELAP5/MOD2 (cycle 36.04) to assess its capabilities in predicting the reflood phenomena. The code has demonstrated the capability to predict experimental behavior; however, refinements in the interfacial drag are required to improve the code's predictions.
- Research Organization:
- Texas A and M Univ., College Station (USA)
- OSTI ID:
- 7118620
- Report Number(s):
- CONF-8711195-
- Conference Information:
- Journal Name: Trans. Am. Nucl. Soc.; (United States) Journal Volume: 55
- Country of Publication:
- United States
- Language:
- English
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Related Subjects
21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS
210200 -- Power Reactors
Nonbreeding
Light-Water Moderated
Nonboiling Water Cooled
22 GENERAL STUDIES OF NUCLEAR REACTORS
220900* -- Nuclear Reactor Technology-- Reactor Safety
ACCIDENTS
AFTER-HEAT REMOVAL
BLOWDOWN
COMPUTER CODES
DATA
ECCS
ENERGY TRANSFER
ENGINEERED SAFETY SYSTEMS
EXPERIMENTAL DATA
FLUID MECHANICS
HEAT TRANSFER
HYDRAULICS
INFORMATION
LOSS OF COOLANT
MECHANICS
NATIONAL ORGANIZATIONS
NUMERICAL DATA
ORNL
PWR TYPE REACTORS
R CODES
REACTOR ACCIDENTS
REACTOR PROTECTION SYSTEMS
REACTOR SAFETY EXPERIMENTS
REACTORS
REMOVAL
TEST FACILITIES
US AEC
US DOE
US ERDA
US ORGANIZATIONS
WATER COOLED REACTORS
WATER MODERATED REACTORS
210200 -- Power Reactors
Nonbreeding
Light-Water Moderated
Nonboiling Water Cooled
22 GENERAL STUDIES OF NUCLEAR REACTORS
220900* -- Nuclear Reactor Technology-- Reactor Safety
ACCIDENTS
AFTER-HEAT REMOVAL
BLOWDOWN
COMPUTER CODES
DATA
ECCS
ENERGY TRANSFER
ENGINEERED SAFETY SYSTEMS
EXPERIMENTAL DATA
FLUID MECHANICS
HEAT TRANSFER
HYDRAULICS
INFORMATION
LOSS OF COOLANT
MECHANICS
NATIONAL ORGANIZATIONS
NUMERICAL DATA
ORNL
PWR TYPE REACTORS
R CODES
REACTOR ACCIDENTS
REACTOR PROTECTION SYSTEMS
REACTOR SAFETY EXPERIMENTS
REACTORS
REMOVAL
TEST FACILITIES
US AEC
US DOE
US ERDA
US ORGANIZATIONS
WATER COOLED REACTORS
WATER MODERATED REACTORS