Assessment of oxygen diffusion during UO/sub 2/--Zircaloy interaction. [PWR]

Technical Report · Sun Jan 01 04:00:00 EST 1978

OSTI ID:5190231

Cronenberg, A W; El-Genk, M S

Off-normal light water reactor fuel element behavior is being studied as part of the Thermal Fuels Behavior Program (TFBP) conducted by EG and G Idaho, Inc., for the Nuclear Regulatory Commission. As part of this program, a series of single- and multiple-rod pressurized water reactor fuel behavior experiments have been performed in the Power Burst Facility (PBF) under various power-cooling-mismatch (PCM) conditions. Although the primary purpose of such integral experiments is to obtain fuel rod thermal and mechanical performance data, several chemical and metallurgical phenomena have been observed as a consequence of testing fuel rods under film boiling conditions. Such phenomena include zircaloy oxidation as a result of both steam-cladding and UO/sub 2/-cladding interaction. Although analytical modeling of the steam-zircaloy interaction process has been developed over the past 15 years, the fuel-cladding interaction process has yet to be satisfactorily modeled.

• Research Organization: Idaho National Engineering Lab., Idaho Falls (USA)
• DOE Contract Number: EY-76-C-07-1570
• OSTI ID: 5190231
• Report Number(s): TREE-NUREG-1192
• Country of Publication: United States
• Language: English

Related Subjects

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS
210200* -- Power Reactors
Nonbreeding
Light-Water Moderated
Nonboiling Water Cooled
ACCIDENTS
ACTINIDE COMPOUNDS
ALLOYS
CHALCOGENIDES
CHEMICAL REACTIONS
FUEL ELEMENTS
FUEL-CLADDING INTERACTIONS
OXIDATION
OXIDES
OXYGEN COMPOUNDS
POWER-COOLING-MISMATCH ACCIDENTS
PWR TYPE REACTORS
REACTOR ACCIDENTS
REACTOR COMPONENTS
REACTORS
TIN ALLOYS
URANIUM COMPOUNDS
URANIUM DIOXIDE
URANIUM OXIDES
WATER COOLED REACTORS
WATER MODERATED REACTORS
ZIRCALOY
ZIRCONIUM ALLOYS
ZIRCONIUM BASE ALLOYS