Semiscale steam-generator tube-rupture test results. [PWR]
Conference
·
OSTI ID:5563759
The Semiscale Program and Test facility are located at the Idaho National Engineering Laboratory, and operated by EG and G Idaho, Inc., for the US Department of Energy. The system is a small-scale model of the primary coolant system of a pressurized water reactor (PWR) nuclear generating plant. An experimental program designed to provide data from steam generator tube rupture is presently being conducted in the Semiscale Mod-2B System in support of programs sponsored by the US Nuclear Regulatory Commission Office of Nuclear Regulatory Research. The test series will provide computer code verification data for a spectrum of tube rupture events including various break sizes (modeling number of tubes ruptured) and various operator safety actions in response to the rupture. The test series will also be used to characterize accident signatures and plant response to typical emergency recovery procedures.
- Research Organization:
- EG and G Idaho, Inc., Idaho Falls (USA)
- DOE Contract Number:
- AC07-76ID01570
- OSTI ID:
- 5563759
- Report Number(s):
- EGG-M-22183; CONF-8310143-37; ON: DE84002204
- Country of Publication:
- United States
- Language:
- English
Similar Records
Semiscale loss-of-power test results. [PWR]
Results of the Semiscale Mod-2B steam generator tube rupture test series
Experiment data report for Semiscale Mod-1 test S-28-3 (steam generator tube rupture test). [PWR]
Technical Report
·
Fri Dec 31 23:00:00 EST 1982
·
OSTI ID:5691875
Results of the Semiscale Mod-2B steam generator tube rupture test series
Technical Report
·
Mon Dec 31 23:00:00 EST 1984
·
OSTI ID:6078644
Experiment data report for Semiscale Mod-1 test S-28-3 (steam generator tube rupture test). [PWR]
Technical Report
·
Sat Oct 01 00:00:00 EDT 1977
·
OSTI ID:5309950
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
BOILERS
DATA
ENERGY TRANSFER
EXPERIMENTAL DATA
FAILURES
FLOW RATE
FLUID MECHANICS
HEAT TRANSFER
HYDRAULICS
INFORMATION
MECHANICS
NUMERICAL DATA
PRESSURE GRADIENTS
PWR TYPE REACTORS
REACTOR ACCIDENTS
REACTOR SAFETY
REACTORS
SAFETY
STEAM GENERATORS
TEST FACILITIES
THEORETICAL DATA
TUBES
VAPOR GENERATORS
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
BOILERS
DATA
ENERGY TRANSFER
EXPERIMENTAL DATA
FAILURES
FLOW RATE
FLUID MECHANICS
HEAT TRANSFER
HYDRAULICS
INFORMATION
MECHANICS
NUMERICAL DATA
PRESSURE GRADIENTS
PWR TYPE REACTORS
REACTOR ACCIDENTS
REACTOR SAFETY
REACTORS
SAFETY
STEAM GENERATORS
TEST FACILITIES
THEORETICAL DATA
TUBES
VAPOR GENERATORS
WATER COOLED REACTORS
WATER MODERATED REACTORS