Pretest analysis document for Test S-FS-6

Technical Report · Wed May 01 00:00:00 EDT 1985

OSTI ID:5697795

Shaw, R A; Hall, D G

This report documents the pretest analyses completed for Semiscale Test S-FS-6. This test will simulate a transient initiated by a 100% break in a steam generator bottom feedwater line downstream of the check valve. The initial conditions represent normal operating conditions for a C-E System 80 nuclear power plant. Predictions of transients resulting from feedwater line breaks in these plants have indicated that significant primary system overpressurization may occur. The enclosed analyses include a RELAP5/MOD2/CY21 code calculation and preliminary results from a facility hot, integrated test which was conducted to near S-FS-6 specifications. The results of these analyses indicate that the test objectives for Test S-FS-6 can be achieved. The primary system overpressurization will pose no threat to personnel or plant integrity.

• Research Organization: EG and G Idaho, Inc., Idaho Falls (USA)
• DOE Contract Number: AC07-76ID01570
• OSTI ID: 5697795
• Report Number(s): EGG-SEMI-6878; ON: TI85014608
• Country of Publication: United States
• Language: English

Related Subjects

22 GENERAL STUDIES OF NUCLEAR REACTORS
21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS
PWR TYPE REACTORS
TRANSIENTS
SIMULATION
FEEDWATER
STEAM GENERATORS
TEST FACILITIES
BOILERS
HYDROGEN COMPOUNDS
OXYGEN COMPOUNDS
REACTORS
VAPOR GENERATORS
WATER
WATER COOLED REACTORS
WATER MODERATED REACTORS
220900* - Nuclear Reactor Technology- Reactor Safety
210200 - Power Reactors
Nonbreeding
Light-Water Moderated
Nonboiling Water Cooled