Late-stage accident behavior of a highly disrupted LMR oxide fuel bundle

Conference · Wed Jan 01 00:00:00 EST 1992 · Transactions of the American Nuclear Society; (United States)

OSTI ID:6681383

Bauer, T H; Holland, J W; Wright, A E ^[1]

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1. Argonne National Lab., IL (United States)

In the prototype fast reactor (PFR) Transient Reactor Test Facility (TREAT) test LO3 performed in 1982, a seven-pin bundle of prototypical 4 at.% burnup PFR oxide fuel pins was heated to failure and beyond in a simulation of a slow ramp rate overpower transient in a liquid-metal-cooled reactor (LMR). Prototypical aspects of the test also included flowing sodium coolant and nuclear heating in the TREAT reactor. The focus of the test was the initial phase of material disruption following initial cladding failure. The LO3 power transient did not terminate shortly after failure as planned, however, but inadvertently continued for [approx] 3.67 s past pin failure until the control rods were fully withdrawn. This unintended extension of the power transient subjected damaged, uncooled fuel to an additional [approx]10 fullpower seconds of heating. Extensive dissolution, reconfiguration, and ultimately relocation of the damaged fuel bundle ensued. This paper describes these previously unreported late-stage postfailure events and draws conclusions regarding fuel behavior.

• OSTI ID: 6681383
• Report Number(s): CONF-921102-; CODEN: TANSAO
• Journal Information: Transactions of the American Nuclear Society; (United States), Vol. 66; Conference: Joint American Nuclear Society (ANS)/European Nuclear Society (ENS) international meeting on fifty years of controlled nuclear chain reaction: past, present, and future, Chicago, IL (United States), 15-20 Nov 1992; ISSN 0003-018X
• Country of Publication: United States
• Language: English

Related Subjects

22 GENERAL STUDIES OF NUCLEAR REACTORS
21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS
FUEL PINS
FAILURES
LIQUID METAL COOLED REACTORS
TRANSIENT OVERPOWER ACCIDENTS
FUEL ELEMENT FAILURE
FUEL-COOLANT INTERACTIONS
MELTING
NUCLEAR FUELS
OXIDES
PFR REACTOR
TESTING
THERMAL ANALYSIS
TREAT REACTOR
ACCIDENTS
AIR COOLED REACTORS
BREEDER REACTORS
CHALCOGENIDES
ENERGY SOURCES
ENRICHED URANIUM REACTORS
EPITHERMAL REACTORS
EXPERIMENTAL REACTORS
FAST REACTORS
FBR TYPE REACTORS
FUEL ELEMENTS
FUELS
GAS COOLED REACTORS
GRAPHITE MODERATED REACTORS
HOMOGENEOUS REACTORS
MATERIALS
OXYGEN COMPOUNDS
PHASE TRANSFORMATIONS
POWER REACTORS
REACTOR ACCIDENTS
REACTOR COMPONENTS
REACTOR MATERIALS
REACTORS
RESEARCH AND TEST REACTORS
SODIUM COOLED REACTORS
SOLID HOMOGENEOUS REACTORS
TEST REACTORS
THERMAL REACTORS
220900* - Nuclear Reactor Technology- Reactor Safety
210500 - Power Reactors
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