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

Conference · Wed Jan 01 00:00:00 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-; TRN: 87-031927
• Journal Information: Trans. Am. Nucl. Soc.; (United States), Vol. 53; Conference: American Nuclear Society and Atomic Industrial Forum joint meeting, Washington, DC, USA, 16 Nov 1986
• Country of Publication: United States
• Language: English

Related Subjects

22 GENERAL STUDIES OF NUCLEAR REACTORS
21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS
FUEL ELEMENTS
PERFORMANCE
LMFBR TYPE REACTORS
POWER-COOLING-MISMATCH ACCIDENTS
REACTOR SAFETY
TREAT REACTOR
REACTOR SAFETY EXPERIMENTS
BURNUP
EBR-2 REACTOR
FAILURES
FUEL CANS
FUEL ELEMENT FAILURE
FUEL PINS
PLUTONIUM
POWER DENSITY
URANIUM
ZIRCONIUM
ACCIDENTS
ACTINIDES
AIR COOLED REACTORS
BREEDER REACTORS
ELEMENTS
ENRICHED URANIUM REACTORS
EPITHERMAL REACTORS
EXPERIMENTAL REACTORS
FAST REACTORS
FBR TYPE REACTORS
GAS COOLED REACTORS
GRAPHITE MODERATED REACTORS
HOMOGENEOUS REACTORS
LIQUID METAL COOLED REACTORS
METALS
POWER REACTORS
REACTOR ACCIDENTS
REACTOR COMPONENTS
REACTORS
RESEARCH AND TEST REACTORS
SAFETY
SODIUM COOLED REACTORS
SOLID HOMOGENEOUS REACTORS
TEST REACTORS
THERMAL REACTORS
TRANSITION ELEMENTS
TRANSURANIUM ELEMENTS
220900* - Nuclear Reactor Technology- Reactor Safety
210500 - Power Reactors
Breeding
220600 - Nuclear Reactor Technology- Research
Test & Experimental Reactors