Thermal-hydraulic effects of clad swelling and rupture during reflood
During a loss-of-coolant accident (LOCA) in a pressurized water reactor (PWR), the cladding of some fuel rods may undergo swelling and rupture. Droplet breakup at swelled and ruptured locations plays an important role in the coolability of these rods during the LOCA reflooding period. Various reflood experiments simulated blockage and rupture effects using sleeves mounted on solid electrically heated rods. The REBEKA tests, which used electrically heated rods with pressurized gaps, show that the formation and propagation of a second quench front from the rupture location are dominant in reducing the clad temperature downstream of the rupture compared to other blockage and rupture effects. The RELAP5/MOD2 computer code was extensively modified by the Babcock Wilcox Fuel Company to improve its predictive capabilities. Recently, models were added to predict the thermal-hydraulic behavior of a nuclear fuel rod during the reflood phase of a large-break LOCA in a PWR. Modifications included modeling to simulate grid spacer, blockage, and rupture effects; a dynamic gap model to calculate gap heat transfer and to predict clad swelling and rupture; and a metal/water reaction model. A simple model to simulate the droplet breakup mechanisms at a grid spacer or clad rupture location was developed and incorporated into RELAP5.
- OSTI ID:
- 6005434
- Report Number(s):
- CONF-901101-; CODEN: TANSA
- Journal Information:
- Transactions of the American Nuclear Society; (USA), Vol. 62; Conference: American Nuclear Society (ANS) winter meeting, Washington, DC (USA), 11-15 Nov 1990; ISSN 0003-018X
- Country of Publication:
- United States
- Language:
- English
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Related Subjects
21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS
FUEL ELEMENTS
QUENCHING
LOSS OF COOLANT
HEAT TRANSFER
HYDRAULICS
PWR TYPE REACTORS
FUEL ELEMENT FAILURE
REACTOR SAFETY
DROPLETS
EFFICIENCY
PIPES
R CODES
REACTOR SAFETY EXPERIMENTS
RUPTURES
SWELLING
TEST FACILITIES
ACCIDENTS
COMPUTER CODES
ENERGY TRANSFER
FAILURES
FLUID MECHANICS
MECHANICS
PARTICLES
REACTOR ACCIDENTS
REACTOR COMPONENTS
REACTORS
SAFETY
WATER COOLED REACTORS
WATER MODERATED REACTORS
220900* - Nuclear Reactor Technology- Reactor Safety
210200 - Power Reactors
Nonbreeding
Light-Water Moderated
Nonboiling Water Cooled