Analysis of in-pile heat transfer tests: Final report

Technical Report · Sat Nov 01 00:00:00 EST 1986

OSTI ID:7037594

Rowe, D S; Nielsen, L H; Henke, R K

This report presents the results of analysis of selected data from the NRU test series dealing with heatup and reflood heat transfer during postulated PWR LOCA conditions. These tests used nuclear fuel rods and some considered clad ballooning and rupture. Also included was an electrically-heated rod ballooning test, REBEKA-6. The COBRA-TF computer program, renamed PYTHONS, was modified and used for the analytical tool. Modifications included provisions for fuel rod gas flow and pressure, creep deformation and rupture, channel blockage, and blockage heat transfer. Calculated clad temperatures for NRU unpressurized rods show quite good agreement with experimental data. The calculated amount and axial extent of clad ballooning for pressurized rods agrees reasonably well with post-test examinations of the NRU bundles. Time to failure was underpredicted in the MT-3 test as a result of the high strength of NRU clad material which was not represented in the PYTHONS creep strain correlation. Use of an alternate creep model in the REBEKA-6 analysis showed excellent prediction of time to failure. Fine detail in the deformation profile caused by the cooling effects of grid spacers was not calculated by the code. Further analytical development would be needed to include this effect.

• Research Organization: Rowe and Associates, Bellevue, WA (USA)
• OSTI ID: 7037594
• Report Number(s): EPRI-NP-4899; ON: TI87920141
• 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
TWO-PHASE FLOW
PWR TYPE REACTORS
REACTOR CORES
TEMPERATURE DISTRIBUTION
CALCULATION METHODS
DEFORMATION
FUEL RODS
HYDRAULICS
NUCLEAR FUELS
P CODES
REACTOR SAFETY
TESTING
ACCIDENTS
COMPUTER CODES
ENERGY SOURCES
ENERGY TRANSFER
FLUID FLOW
FLUID MECHANICS
FUEL ELEMENTS
FUELS
MATERIALS
MECHANICS
REACTOR ACCIDENTS
REACTOR COMPONENTS
REACTOR MATERIALS
REACTORS
SAFETY
WATER COOLED REACTORS
WATER MODERATED REACTORS
220900* - Nuclear Reactor Technology- Reactor Safety
210200 - Power Reactors
Nonbreeding
Light-Water Moderated
Nonboiling Water Cooled