Pretest analysis document for Test S-FS-11

Technical Report · Mon Jul 01 00:00:00 EDT 1985

OSTI ID:5278417

Hall, D G; Shaw, R A

This report documents the pretest calculations completed for Semiscale Test S-FS-11. This test will simulate a transient initiated by a 50% break in a steam generator bottom feedwater line downstream of the check valve. The initial conditions represents normal operating conditions for a C-E System 80 nuclear plant. Prediction of transients resulting from feedwater line breaks in these plants have indicated that significant primary system overpressurization may occur. The results of a RELAP5/MOD2/CY21 code calculation indicate that the test objectives for Test S-FS-11 can be achieved. The primary system overpressurization will occur but pose no threat to personnel or plant integrity. 3 refs., 15 figs., 5 tabs.

• Research Organization: EG and G Idaho, Inc., Idaho Falls (USA)
• DOE Contract Number: AC07-76ID01570
• OSTI ID: 5278417
• Report Number(s): EGG-SEMI-6985; ON: TI85017495
• Country of Publication: United States
• Language: English

Related Subjects

22 GENERAL STUDIES OF NUCLEAR REACTORS
21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS
CE STANDARD REACTOR
LOSS OF COOLANT
PRIMARY COOLANT CIRCUITS
HEAT TRANSFER
HYDRAULICS
COMPUTERIZED SIMULATION
PRESSURIZING
STEAM GENERATORS
ACCIDENTS
BOILERS
COOLING SYSTEMS
ENERGY SYSTEMS
ENERGY TRANSFER
FLUID MECHANICS
MECHANICS
PWR TYPE REACTORS
REACTOR ACCIDENTS
REACTOR COMPONENTS
REACTOR COOLING SYSTEMS
REACTORS
SIMULATION
VAPOR GENERATORS
WATER COOLED REACTORS
WATER MODERATED REACTORS
220900* - Nuclear Reactor Technology- Reactor Safety
210200 - Power Reactors
Nonbreeding
Light-Water Moderated
Nonboiling Water Cooled