Experiment data report for LOFT nonnuclear Test L1-4. [PWR]
Test L1-4 was the fourth in a series of five nonnuclear isothermal blowdown tests conducted by the Loss of Fluid Test (LOFT) Program. Test L1-4 was the first Nuclear Regulatory Commission standard problem (International Problem No. 5 and U.S. Problem No. 7) experiment conducted at LOFT. Data from this test will be compared with predictions generated by the standard problem participants. For this test the LOFT Facility was configured to simulate a loss-of-coolant accident in a large pressurized water reactor resulting from a 200% double-ended offset shear break in a cold leg of the primary coolant system. A hydraulic core simulator assembly was installed in place of the nuclear core. The initial conditions in the primary coolant system intact loop were temperature at 279/sup 0/C, gauge pressure at 15.65 MPa, and intact loop flow at 268.4 kg/s. During system depressurization into a simulated containment, emergency core cooling water was injected into the primary coolant system cold leg to provide data on the effects of emergency core cooling on system thermalhydraulic response.
- Research Organization:
- EG and G Idaho, Inc., Idaho Falls (USA)
- Sponsoring Organization:
- US Energy Research and Development Administration (ERDA)
- DOE Contract Number:
- EY-76-C-07-1570
- OSTI ID:
- 7088464
- Report Number(s):
- TREE-NUREG-1084
- Country of Publication:
- United States
- Language:
- English
Similar Records
Experiment data report for LOFT nonnuclear Test L1-5 (isothermal test with Core 1 installed)
Experiment data report for LOFT nonnuclear Test L1-1
Experimental data report for LOFT nonnuclear test L1-3A
Technical Report
·
Thu Jun 01 00:00:00 EDT 1978
·
OSTI ID:6863061
Experiment data report for LOFT nonnuclear Test L1-1
Technical Report
·
Fri Dec 31 23:00:00 EST 1976
·
OSTI ID:7225028
Experimental data report for LOFT nonnuclear test L1-3A
Technical Report
·
Tue Nov 30 23:00:00 EST 1976
·
OSTI ID:7225027
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
BLOWDOWN
ECCS
ENERGY TRANSFER
ENGINEERED SAFETY SYSTEMS
FLOW RATE
FLUID FLOW
HEAT TRANSFER
HYDRAULICS
LIQUID FLOW
LOSS OF COOLANT
MOCKUP
PRESSURE GRADIENTS
PWR TYPE REACTORS
REACTOR ACCIDENTS
REACTOR PROTECTION SYSTEMS
REACTOR SAFETY
REACTORS
SAFETY
STRUCTURAL MODELS
TEMPERATURE GRADIENTS
TEST FACILITIES
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
BLOWDOWN
ECCS
ENERGY TRANSFER
ENGINEERED SAFETY SYSTEMS
FLOW RATE
FLUID FLOW
HEAT TRANSFER
HYDRAULICS
LIQUID FLOW
LOSS OF COOLANT
MOCKUP
PRESSURE GRADIENTS
PWR TYPE REACTORS
REACTOR ACCIDENTS
REACTOR PROTECTION SYSTEMS
REACTOR SAFETY
REACTORS
SAFETY
STRUCTURAL MODELS
TEMPERATURE GRADIENTS
TEST FACILITIES
WATER COOLED REACTORS
WATER MODERATED REACTORS