Effect of cladding defect size on the oxidation of irradiated spent LWR (light-water reactor) fuel below 369/sup 0/C

Conference · Wed Oct 17 00:00:00 EDT 1984

OSTI ID:6502023

Einziger, R E; Strain, R V

Tests on spent fuel fragments and rod segments were conducted between 250 and 360/sup 0/C to relate temperature, defect size, and fuel oxidation rate with time-to-cladding-splitting. Defect sizes from 760 ..mu..m diameter down to 8 ..mu..m, the size of an SCC type breach, were used. Above 283/sup 0/C, the time-to-cladding-splitting was longer for the smaller defects. The enhancement of the incubation time by smaller defects steadily decreased with temperature and was not detected at 250/sup 0/C. 18 refs., 10 figs., 4 tabs.

• Research Organization: Hanford Engineering Development Lab., Richland, WA (USA); Argonne National Lab., IL (USA)
• DOE Contract Number: AC06-76FF02170
• OSTI ID: 6502023
• Report Number(s): HEDL-SA-3168-FP; CONF-841094-6; ON: DE87008383
• Country of Publication: United States
• Language: English

Related Subjects

050700 -- Nuclear Fuels-- Fuels Production & Properties
050900* -- Nuclear Fuels-- Transport
Handling
& Storage
11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS
ACTINIDE COMPOUNDS
CHALCOGENIDES
CHEMICAL REACTIONS
DEFECTS
DRY STORAGE
ENERGY SOURCES
FAILURES
FUEL CANS
FUELS
MATERIALS
NUCLEAR FUELS
OXIDATION
OXIDES
OXYGEN COMPOUNDS
REACTOR MATERIALS
SPENT FUEL STORAGE
SPENT FUELS
STORAGE
URANIUM COMPOUNDS
URANIUM DIOXIDE
URANIUM OXIDES