Effect of reflood prediction uncertainties on LOFT cladding oxidation
Conference
·
OSTI ID:6609422
The FLOOD4 and RELAP4/MOD6 computer codes, which are used to perform LOFT reflood analysis, have been compared to FLECHT-SET and Semiscale gravity feed tests to provide an evaluation of core reflood prediction techniques and an identification of phenomena important to LOFT reflood behavior. These comparisons provide a basis for estimating uncertainty in cladding temperature history during the LOFT loss-of-coolant experiments (LOCEs). The bounds on the cladding temperature response are then utilized to estimate a range of expected cladding oxidation and embrittlement which is essential for identifying special equipment needed during replacement, storage, and post-test examination of LOFT fuel modules. FLOOD4 couples the system hydraulic response with core heat transfer and steam generation. Four heat transfer correlations simulate the boiling curve and liquid entrainment, fallback and vaporization in the steam generators are modeled. FLOOD4 requires user input multipliers to specify the dispersed flow heat transfer, liquid entrainment and correlations to describe liquid fallback from the upper plenum region. The fraction of liquid vaporized in the steam generators must also be user input.
- Research Organization:
- EG and G Idaho, Inc., Idaho Falls (USA)
- DOE Contract Number:
- EY-76-C-07-1570
- OSTI ID:
- 6609422
- Report Number(s):
- CONF-781105-34
- Country of Publication:
- United States
- Language:
- English
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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
CHEMICAL REACTIONS
COMPUTER CALCULATIONS
COMPUTER CODES
ECCS
ENGINEERED SAFETY SYSTEMS
F CODES
FUEL CANS
LOFT REACTOR
LOSS OF COOLANT
OXIDATION
PERFORMANCE TESTING
PWR TYPE REACTORS
R CODES
REACTOR ACCIDENTS
REACTOR PROTECTION SYSTEMS
REACTORS
RESEARCH AND TEST REACTORS
TANK TYPE REACTORS
TEMPERATURE DISTRIBUTION
TEST REACTORS
TESTING
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
CHEMICAL REACTIONS
COMPUTER CALCULATIONS
COMPUTER CODES
ECCS
ENGINEERED SAFETY SYSTEMS
F CODES
FUEL CANS
LOFT REACTOR
LOSS OF COOLANT
OXIDATION
PERFORMANCE TESTING
PWR TYPE REACTORS
R CODES
REACTOR ACCIDENTS
REACTOR PROTECTION SYSTEMS
REACTORS
RESEARCH AND TEST REACTORS
TANK TYPE REACTORS
TEMPERATURE DISTRIBUTION
TEST REACTORS
TESTING
WATER COOLED REACTORS
WATER MODERATED REACTORS