Behavior of metallic uranium-fissium fuel in TREAT transient overpower tests

Conference · Wed Jan 01 04:00:00 EST 1986 · Trans. Am. Nucl. Soc.; (United States)

OSTI ID:6298702

Bauer, T H; Klickman, A E; Lo, R K; Rhodes, E A; Robinson, W R; Stanford, G S; Wright, A E

TREAT tests M2, M3, and M4 were performed to obtain information on two key behavior characteristics of fuel under transient overpower accident conditions in metal-fueled fast reactors: the prefailure axial self-extrusion (elongation beyond thermal expansion) of fuel within intact cladding and the margin to cladding breach. Uranium-5 wt% fissium Experimental Breeder Reactor-II driver fuel pins were used for the tests since they were available as suitable stand-ins for the uranium-plutonium-zirconium ternary fuel, which is the reference fuel of the integral fast reactor (IFR) concept. The ternary fuel will be used in subsequent TREAT tests. Preliminary results from tests M2 and M3 were presented earlier. The present report includes significant advances in analysis as well as additional data from test M4. Test results and analysis have led to the development and validation of pin cladding failure and fuel extrusion models for metallic fuel, within reasonable uncertainties for the uranium-fissium alloy. Concepts involved are straightforward and readily extendable to ternary alloys and behavior in full-size reactors.

• Research Organization: Argonne National Lab., IL
• OSTI ID: 6298702
• Report Number(s): CONF-861102-
• Conference Information: Journal Name: Trans. Am. Nucl. Soc.; (United States) Journal Volume: 53
• Country of Publication: United States
• Language: English

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ACCIDENTS
ACTINIDES
AIR COOLED REACTORS
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LIQUID METAL COOLED REACTORS
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METALS
PERFORMANCE
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POWER-COOLING-MISMATCH ACCIDENTS
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REACTORS
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SOLID HOMOGENEOUS REACTORS
TEST REACTORS
THERMAL REACTORS
TRANSITION ELEMENTS
TRANSURANIUM ELEMENTS
TREAT REACTOR
URANIUM
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