Thermal-hydraulic effects of clad swelling and rupture during reflood

Conference · Mon Jan 01 04:00:00 EST 1990 · Transactions of the American Nuclear Society; (USA)

OSTI ID:6005434

Nithianandan, C K; Biller, J R

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--
• Conference Information: Journal Name: Transactions of the American Nuclear Society; (USA) Journal Volume: 62
• Country of Publication: United States
• Language: English

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ACCIDENTS
COMPUTER CODES
DROPLETS
EFFICIENCY
ENERGY TRANSFER
FAILURES
FLUID MECHANICS
FUEL ELEMENT FAILURE
FUEL ELEMENTS
HEAT TRANSFER
HYDRAULICS
LOSS OF COOLANT
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QUENCHING
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REACTOR ACCIDENTS
REACTOR COMPONENTS
REACTOR SAFETY
REACTOR SAFETY EXPERIMENTS
REACTORS
RUPTURES
SAFETY
SWELLING
TEST FACILITIES
WATER COOLED REACTORS
WATER MODERATED REACTORS