In-pile measurement of the thermal conductivity of irradiated metallic fuel

Conference · Wed Jan 01 04:00:00 EST 1992 · Transactions of the American Nuclear Society; (United States)

OSTI ID:7056042

Bauer, T H; Holland, J W ^[1]

---

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

Reliable knowledge of thermal conductivity is needed to assess reactor fuel performance. In normal operation, fuel pin linear power and fuel thermal conductivity together largely determine the fuel's peak temperature, and under severe accident conditions, fuel thermal conductivity largely determines the power to melt. Unfortunately, assessing the thermal conductivity of modern high-swelling sodium-bonded metallic fuel is neither theoretically straightforward nor amenable to laboratory measurements. At issue is whether the high swelling that takes place in metallic fuel during the first few atomic percent of burnup could lead to drastic conductivity reductions by factors of as much as 0.4 to 0.5. Such reduction may be mitigated by the infiltration of fuel porosity by high-conductivity liquid bond sodium or later by the solid and liquid fission products that accumulate with burnup. In this paper, the authors utilize measurements of maximum melting from intact irradiated metal fuel pins tested in the recent M-series in-pile test program in TREAT to estimate metal fuel thermal conductivity directly and in situ. Of six irradiated integral fast reactor prototype fuel pins ubjected to severe overpower and melting, four pins remained intact. Three of the test pins were of ternary alloy (U-19 Pu-10 Zr) and one of binary (U-10 Zr) alloy. Importantly, these test pins span key range of low burnup where minimum values of conductivity would be expected.

• OSTI ID: 7056042
• Report Number(s): CONF-920606--
• Conference Information: Journal Name: Transactions of the American Nuclear Society; (United States) Journal Volume: 65
• Country of Publication: United States
• Language: English

Related Subjects

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS
210500 -- Power Reactors
Breeding
22 GENERAL STUDIES OF NUCLEAR REACTORS
220900* -- Nuclear Reactor Technology-- Reactor Safety
ACCIDENTS
ACTINIDES
AIR COOLED REACTORS
ALLOY NUCLEAR FUELS
BURNUP
ELEMENTS
ENERGY SOURCES
ENERGY TRANSFER
ENRICHED URANIUM REACTORS
EPITHERMAL REACTORS
EXPERIMENTAL REACTORS
FAST REACTORS
FLUID MECHANICS
FUEL ELEMENTS
FUEL INTEGRITY
FUELS
GAS COOLED REACTORS
GRAPHITE MODERATED REACTORS
HEAT TRANSFER
HOMOGENEOUS REACTORS
HYDRAULICS
IFR REACTOR
IRRADIATION
MATERIALS
MECHANICS
MELTING
METALS
NUCLEAR FUELS
PERFORMANCE
PHASE TRANSFORMATIONS
PHYSICAL PROPERTIES
PLUTONIUM
REACTOR ACCIDENTS
REACTOR COMPONENTS
REACTOR MATERIALS
REACTOR SAFETY
REACTORS
RELIABILITY
RESEARCH AND TEST REACTORS
SAFETY
SOLID FUELS
SOLID HOMOGENEOUS REACTORS
SWELLING
TEST REACTORS
THERMAL CONDUCTIVITY
THERMAL REACTORS
THERMODYNAMIC PROPERTIES
TRANSIENT OVERPOWER ACCIDENTS
TRANSITION ELEMENTS
TRANSURANIUM ELEMENTS
TREAT REACTOR
ZERO POWER REACTORS
ZIRCONIUM