Experiment data report for semiscale Mod-3 integral blowdown and reflood heat transfer test S-07-6 (Baseline Test Series). [PWR]
Technical Report
·
OSTI ID:6399863
Recorded test data are presented for Test S-07-6 of the Semiscale Mod-3 baseline test series. This test is one of several Semiscale Mod-3 experiments conducted to investigate the thermal and hydraulic phenomena accompanying a hypothesized loss-of-coolant accident in a pressurized water reactor (PWR) system. Test S-07-6 was conducted from initial conditions of 15.21 MPa and 559 K to investigate the response of the Semiscale Mod-3 system to a blowdown transient following a simulated double-ended offset shear of the broken loop cold leg piping. The specific objective of this test was to provide reference data to evaluate integral blowdown and reflood behavior during a 200% cold leg break with emergency core coolant (ECC) injection into the intact loop cold leg of the Mod-3 system. The purpose of this report is to make available the uninterpreted data from Test S-07-6 for future data analysis. The data, presented in the form of graphs in engineering units, have been analyzed only to the extent necessary to ensure that they are reasonable and consistent.
- Research Organization:
- Idaho National Engineering Lab., Idaho Falls (USA)
- DOE Contract Number:
- EY-76-C-07-1570
- OSTI ID:
- 6399863
- Report Number(s):
- NUREG/CR-0467; TREE-1226
- 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
42 ENGINEERING
420400 -- Engineering-- Heat Transfer & Fluid Flow
ACCIDENTS
BLOWDOWN
ECCS
ENERGY TRANSFER
ENGINEERED SAFETY SYSTEMS
FLUID FLOW
FLUID MECHANICS
HEAT TRANSFER
HYDRODYNAMICS
LOSS OF COOLANT
MECHANICS
MOCKUP
PERFORMANCE TESTING
PWR TYPE REACTORS
REACTOR ACCIDENTS
REACTOR PROTECTION SYSTEMS
REACTORS
SIMULATION
STRUCTURAL MODELS
TESTING
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
42 ENGINEERING
420400 -- Engineering-- Heat Transfer & Fluid Flow
ACCIDENTS
BLOWDOWN
ECCS
ENERGY TRANSFER
ENGINEERED SAFETY SYSTEMS
FLUID FLOW
FLUID MECHANICS
HEAT TRANSFER
HYDRODYNAMICS
LOSS OF COOLANT
MECHANICS
MOCKUP
PERFORMANCE TESTING
PWR TYPE REACTORS
REACTOR ACCIDENTS
REACTOR PROTECTION SYSTEMS
REACTORS
SIMULATION
STRUCTURAL MODELS
TESTING
WATER COOLED REACTORS
WATER MODERATED REACTORS