In-pile TREAT Test L04: simulating a lead sub-assembly in an unprotected LMFBR loss-of-coolant accident
Test L04 in the PFR/TREAT series is the first multi-pin, in-pile simulation of a LMFBR transient undercooling/overpower (TUCOP) accident using full length prototypic fuel irradiated in a fast reactor. L04 is a gridded 7-pin bundle test performed in the ANL Mk-III integral loop in a flowing sodium environment and uses prototypic, bottom plenum, UK reactor fuel, preirradiated in the PFR to an axial peak burn-up of 4.2 a/o. The objective of L04 was the study, by simulation, of coolant voiding and fuel motion during the initiating phase of a hypothetical TUCOP accident in a large LMFBR. Test L04 is intended to study the behavior of a centrally located, lead subassembly with the highest power-to-flow ratio.
- Research Organization:
- Argonne National Lab., IL (USA)
- DOE Contract Number:
- W-31109-ENG-38
- OSTI ID:
- 5701417
- Report Number(s):
- CONF-831047-30; ON: DE83014678
- Country of Publication:
- United States
- Language:
- English
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Conference
·
Sat Dec 31 23:00:00 EST 1983
·
OSTI ID:6339047
PFR/TREAT Tests L04 and L06: irradiated versus fresh LMFBR fuel under TUCOP accident conditions
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Sat Dec 31 23:00:00 EST 1983
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OSTI ID:6065226
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Technical Report
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Sat Dec 31 23:00:00 EST 1983
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OSTI ID:7143046
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
AIR COOLED REACTORS
BREEDER REACTORS
DATA
ENERGY TRANSFER
ENRICHED URANIUM REACTORS
EPITHERMAL REACTORS
EXPERIMENTAL DATA
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FAST REACTORS
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FLUID MECHANICS
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GRAPHITE MODERATED REACTORS
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IN PILE LOOPS
INFORMATION
LIQUID METAL COOLED REACTORS
LMFBR TYPE REACTORS
LOSS OF COOLANT
MECHANICS
NUMERICAL DATA
PRESSURE GRADIENTS
REACTOR ACCIDENTS
REACTOR COMPONENTS
REACTOR EXPERIMENTAL FACILITIES
REACTOR SAFETY
REACTORS
RESEARCH AND TEST REACTORS
SAFETY
SOLID HOMOGENEOUS REACTORS
TEMPERATURE GRADIENTS
TEST REACTORS
THERMAL REACTORS
TREAT REACTOR
210500 -- Power Reactors
Breeding
22 GENERAL STUDIES OF NUCLEAR REACTORS
220900* -- Nuclear Reactor Technology-- Reactor Safety
ACCIDENTS
AIR COOLED REACTORS
BREEDER REACTORS
DATA
ENERGY TRANSFER
ENRICHED URANIUM REACTORS
EPITHERMAL REACTORS
EXPERIMENTAL DATA
EXPERIMENTAL REACTORS
FAST REACTORS
FBR TYPE REACTORS
FLOW RATE
FLUID MECHANICS
GAS COOLED REACTORS
GRAPHITE MODERATED REACTORS
HEAT TRANSFER
HOMOGENEOUS REACTORS
HYDRAULICS
IN PILE LOOPS
INFORMATION
LIQUID METAL COOLED REACTORS
LMFBR TYPE REACTORS
LOSS OF COOLANT
MECHANICS
NUMERICAL DATA
PRESSURE GRADIENTS
REACTOR ACCIDENTS
REACTOR COMPONENTS
REACTOR EXPERIMENTAL FACILITIES
REACTOR SAFETY
REACTORS
RESEARCH AND TEST REACTORS
SAFETY
SOLID HOMOGENEOUS REACTORS
TEMPERATURE GRADIENTS
TEST REACTORS
THERMAL REACTORS
TREAT REACTOR