FRAP-T6 uncertainty study of LOCA tests LOFT L2-3 and PBF LLR-3. [PWR]

Conference · Sat Jan 01 00:00:00 EST 1983

OSTI ID:6807362

Chambers, R; Driskell, W E; Resch, S C

This paper presents the accuracy and uncertainty of fuel rod behavior calculations performed by the transient Fuel Rod Analysis Program (FRAP-T6) during large break loss-of-coolant accidents. The accuracy of the code was determined primarily through comparisons of code calculations with cladding surface temperature measurements from two loss-of-coolant experiments (LOCEs). These LOCEs were the L2-3 experiment conducted in the Loss-of-Fluid Test (LOFT) Facility and the LOFT Lead Rod 3 (LLR-3) experiment conducted in the Power Burst Facility (PBF). Uncertainties in code calculations resulting from uncertainties in fuel and cladding design variables, material property and heat transfer correlations, and thermal-hydraulic boundary conditions were analyzed.

• Research Organization: EG and G Idaho, Inc., Idaho Falls (USA)
• DOE Contract Number: AC07-76ID01570
• OSTI ID: 6807362
• Report Number(s): EGG-M-08482; CONF-830103-40; ON: DE83005849
• Resource Relation: Conference: 2. international topical meeting on nuclear reactor thermal hydraulics, Santa Barbara, CA, USA, 11 Jan 1983; Other Information: Portions are illegible in microfiche products
• Country of Publication: United States
• Language: English

Related Subjects

22 GENERAL STUDIES OF NUCLEAR REACTORS
21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS
LOSS OF COOLANT
HEAT TRANSFER
HYDRAULICS
PWR TYPE REACTORS
COMPUTER CALCULATIONS
FUEL CANS
FUEL ELEMENT FAILURE
PRESSURE GRADIENTS
REACTOR SAFETY
TEMPERATURE GRADIENTS
ACCIDENTS
ENERGY TRANSFER
FLUID MECHANICS
MECHANICS
REACTOR ACCIDENTS
REACTORS
SAFETY
WATER COOLED REACTORS
WATER MODERATED REACTORS
220900* - Nuclear Reactor Technology- Reactor Safety
210200 - Power Reactors
Nonbreeding
Light-Water Moderated
Nonboiling Water Cooled