Thermal analysis of a six-channel heat-generating blockage in an LMFBR
This paper presents a case study of the temperature fields within and around a six-channel blockage designed as a molten-fuel-release initiator in SLSF-P4, an in-reactor experiment (37-mixed-oxide pin bundle) planned for February, 1981, irradiation. To meet the experiment objectives, a minimum of ten grams of molten UO/sub 2/ must be ejected into the sodium stream from one, two, or three such blockages. The temperature fields of the electrodeposited-nickel blockage filled with a mixture of UO/sub 2/ powder, stainless steel, and gas are found at intervals of full power. The SS content, type of gas, and porosity were parameters varied in this study which used the computer codes THYME-B, SABRE-1, and ANL's version of THTB. State-of-the-art treatments of the conductivity of the mixture and the gas-gap conductance are included. The contrived-blockage design has been found to maintain structural integrity until sufficient molten fuel exists to release, challenge the subassembly, and be detected by delayed-neutron and fission-product monitors. This will serve to resolve lingering questions on rapid pin-to-pin propagation, blockage propagation, and other local-fault issues.
- Research Organization:
- Argonne National Lab., IL (USA)
- DOE Contract Number:
- W-31-109-ENG-38
- OSTI ID:
- 6384655
- Report Number(s):
- CONF-801002-15; ON: DE81023713; TRN: 81-011249
- Resource Relation:
- Conference: ANS/ASME topical meeting on reactor thermal-hydraulics, Saratoga, NY, USA, 9 Oct 1980
- Country of Publication:
- United States
- Language:
- English
Similar Records
Sodium Loop Safety Facility experiment P4. [LMFBR]
SLSF in-reactor local fault safety experiment P4. Final report
Related Subjects
21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS
FLOW BLOCKAGE
HEAT TRANSFER
HYDRAULICS
FUEL-COOLANT INTERACTIONS
REACTOR SAFETY EXPERIMENTS
LMFBR TYPE REACTORS
CORIUM
FUEL ASSEMBLIES
FUEL ELEMENT FAILURE
FUEL PINS
LIQUID FLOW
REACTOR CORE DISRUPTION
REACTOR SAFETY
TEMPERATURE DISTRIBUTION
TEST FACILITIES
ACCIDENTS
BREEDER REACTORS
ENERGY TRANSFER
EPITHERMAL REACTORS
FAST REACTORS
FBR TYPE REACTORS
FLUID FLOW
FLUID MECHANICS
FUEL ELEMENTS
LIQUID METAL COOLED REACTORS
MECHANICS
REACTOR ACCIDENTS
REACTOR COMPONENTS
REACTORS
SAFETY
220900* - Nuclear Reactor Technology- Reactor Safety
210500 - Power Reactors
Breeding