Semiscale Mod-2A Intermediate Break Test series: test-results comparison. [PWR]
Technical Report
·
OSTI ID:6533963
Results are presented from an analysis of Semiscale Mod-2A Intermediate Break Tests S-IB-1, -2, and -3. The tests were 100% (percentage of cold leg pipe flow area), 50%, and 21.7%, respectively, communicative cold leg break loss-of-coolant experiments. They were intended to provide reference data for evaluation and assessment of reactor safety code capabilities to predict integral blowdown, refill/reflood experiments for intermediate break sizes, and, particularly, to provide data to extend the code into the reflood regime. Comparisons of Semiscale intermediate break test results with those from large and small break tests provided characterization of the phenomena observed during the intermediate break tests. An additional objective of Test S-IB-3 was to provide reference data for comparison of Semiscale test results with results from LOFT Test L5-1 and LOBI Test B-R1M.
- Research Organization:
- EG and G Idaho, Inc., Idaho Falls (USA)
- DOE Contract Number:
- AC07-76ID01570
- OSTI ID:
- 6533963
- Report Number(s):
- NUREG/CR-3126; EGG-2238; ON: DE83007847
- Country of Publication:
- United States
- Language:
- English
Similar Records
Experiment data report for Semiscale Mod-2A Intermediate Break Test Series (Test S-IB-3). [Data on microfiche]
Experiment data report for Semiscale Mod-2A Intermediate Break Test Series (Test S-IB-3). [Data on microfiche]
Assessment of RELAP5/MOD2 using semiscale intermediate break loss-of-coolant experiment S-IB-3
Technical Report
·
Tue Jun 01 00:00:00 EDT 1982
·
OSTI ID:5235400
Experiment data report for Semiscale Mod-2A Intermediate Break Test Series (Test S-IB-3). [Data on microfiche]
Technical Report
·
Tue Jun 01 00:00:00 EDT 1982
·
OSTI ID:6947055
Assessment of RELAP5/MOD2 using semiscale intermediate break loss-of-coolant experiment S-IB-3
Technical Report
·
Mon Jun 01 00:00:00 EDT 1992
·
OSTI ID:10156389
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
ENERGY TRANSFER
FLUID MECHANICS
HEAT TRANSFER
HYDRAULICS
LOSS OF COOLANT
MECHANICS
PRESSURE GRADIENTS
PWR TYPE REACTORS
REACTOR ACCIDENTS
REACTOR SAFETY
REACTORS
SAFETY
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
ENERGY TRANSFER
FLUID MECHANICS
HEAT TRANSFER
HYDRAULICS
LOSS OF COOLANT
MECHANICS
PRESSURE GRADIENTS
PWR TYPE REACTORS
REACTOR ACCIDENTS
REACTOR SAFETY
REACTORS
SAFETY
TEMPERATURE GRADIENTS
TEST FACILITIES
WATER COOLED REACTORS
WATER MODERATED REACTORS