TMI-2 upper-core particle bed thermal behavior
Models of dryout heat fluxes of particle beds believed to be applicable to the TMI-2 upper-core particle bed are reviewed and developed. A simplified Lipinski model and a model based on flooding are shown to agree between themselves and with experiments. These models are applied to the calculation of the dryout heat flux of the TMI-2 upper-core particle bed. The TMI-2 upper-core particle bed is shown to be: (a) coolable, if little heat is transferred to it from the consolidated region below, (b) only marginally coolable, if not uncoolable, before material relocation from the consolidated region, if most of the heat in the consolidiated region is transferred to it, and (c) coolable, after the relocation, regardless of heat transfer from the remaining consolidated region. Based on an analogy to quenching experiments, which show that the heat flux during the quench of a particle bed is approximately equal to the dryout heat flux, the time required to quench the TMI-2 upper-core particle bed from 2000 K to the saturation temperature of water during the accident is estimated. The bed was either quenched by 225 min after the initiation of the accident (assuming no heat was transferred to it from the consolidated region) or, at the latest, by 245 min (20 min after molten material relocation to the lower plenum from the consolidated region; assuming most of the heat generated in the consolidated region, both before and after the relocation, was transferred to the particle bed).
- Research Organization:
- EG and G Idaho, Inc., Idaho Falls, ID (United States)
- Sponsoring Organization:
- USDOE, Washington, DC (United States)
- DOE Contract Number:
- AC07-76ID01570
- OSTI ID:
- 10183581
- Report Number(s):
- EGG-TMI-7757; ON: DE92040953
- Resource Relation:
- Other Information: PBD: Aug 1987
- Country of Publication:
- United States
- Language:
- English
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Related Subjects
21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS
THREE MILE ISLAND-2 REACTOR
REACTOR CORES
DRYOUT
REACTOR SAFETY
HEAT TRANSFER
HYDRAULICS
REACTOR COOLING SYSTEMS
LOSS OF COOLANT
REACTOR CORE DISRUPTION
LAMINAR FLOW
TURBULENT FLOW
220900
210200
POWER REACTORS
NONBREEDING
LIGHT-WATER MODERATED
NONBOILING WATER COOLED