FLASH predictions of the MB-2 steam line break tests

Conference · Thu Dec 31 00:00:00 EST 1992

OSTI ID:10106485

Lincoln, F W; Coffield, R D; Johnson, E G

If a main steam line from a pressurized water reactor (PWR) steam generator were to rupture, the effect would be a depressurization of the secondary side and a consequential overcooling transient on the primary side. Analyses must accurately predict the effects of the rapid cooldown of the reactor vessel coolant on positive nuclear-kinetic reactivity feedback to the core plus thermal shock to the reactor vessel and other primary system components. Many early studies of the steam line break (SLB) transient made extremely conservative assumptions to maximize the primary to secondary heat transfer which in turn maximized the reactor vessel cooldown rate. Among the more significant of these assumptions was that flow from the break was pure steam and that the tube bundle remained covered until the secondary mass inventory was significantly reduced. The Model F commercial PWR steam generator testing performed in the Model Boiler No. 2 (MB-2) facility located at the Westinghouse Engineering Test Facility in Tampa, Florida provided data to better qualify the actual variation in these key parameters. A conclusion of this analysis is that the MB-2 steam line break data base is accurate and of sufficient detail to provide a valuable basis for making comparisons relative to code predictions. Results obtained using the FLASH transient safety analysis code were found to be in excellent agreement with the data.

• Research Organization: Westinghouse Electric Corp., West Mifflin, PA (United States). Bettis Atomic Power Lab.
• Sponsoring Organization: USDOE, Washington, DC (United States)
• DOE Contract Number: AC11-93PN38195
• OSTI ID: 10106485
• Report Number(s): WAPD-T-2966; CONF-9211284-1; ON: DE94003725; TRN: 94:001412
• Resource Relation: Conference: American Nuclear Society winter meeting,Washington, DC (United States),Nov 1992; Other Information: PBD: [1992]
• 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
STEAM GENERATORS
RUPTURES
STEAM SYSTEMS
DEPRESSURIZATION
TRANSIENTS
F CODES
REACTOR SAFETY
TEST FACILITIES
HEAT TRANSFER
HYDRAULICS
220900
210200
POWER REACTORS
NONBREEDING
LIGHT-WATER MODERATED
NONBOILING WATER COOLED