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Recent metal fuel safety tests in TREAT

Conference ·
OSTI ID:5737029
; ; ; ;
In-reactor safety tests have been performed on metal-alloy reactor fuel to study its response to transient-overpower conditions, in particular, the margin to cladding breach and the axial self-extrusion of fuel within intact cladding. Uranium-fissium EBR-II driver fuel elements of several burnups were tested, some to cladding breach and others to incipient breach. Transient fuel motions were monitored, and time and location of breach were measured. The test results and computations of fuel extrusion and cladding failure in metal-alloy fuel are described.
Research Organization:
Argonne National Lab., IL (USA)
DOE Contract Number:
W-31109-ENG-38
OSTI ID:
5737029
Report Number(s):
CONF-860501-9; ON: DE86012125
Country of Publication:
United States
Language:
English

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21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS
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AIR COOLED REACTORS
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ENRICHED URANIUM REACTORS
EXPERIMENTAL REACTORS
FAILURES
FUELS
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MATERIALS
NUCLEAR FUELS
REACTOR MATERIALS
REACTOR SAFETY
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RESEARCH AND TEST REACTORS
SAFETY
SOLID HOMOGENEOUS REACTORS
SWELLING
TEST REACTORS
TESTING
THERMAL REACTORS
TRANSIENTS
TREAT REACTOR