Code for predicting the temperature and oxidation of undercooled cores
This report documents the status of a computer code developed by Argonne National Laboratory (ANL) and the Nuclear Safety Analysis Center (NSAC) for predicting temperatures and oxidation of a pressurized-water reactor (open lattice) core during an undercooling transient. The initial use of this code has been for analyzing the TMI-2 core initial uncovery and core heat up during the accident of March, 1979. In the initial investigations of the TMI-2 accident, it was concluded that the first major core damage occurred during the first period of complete loss of forced coolant circulation in the core (100 to 174 minutes after turbine trip). The ANL/NSAC core heatup code was originally created to focus on the TMI-2 core during this uncovery period. This focus greatly simplifies the modeling and computational needs through elimination, as either irrelevant or unresolvable, of specific models for the balance of the primary loop and the secondary loop.
- Research Organization:
- Argonne National Lab., IL (USA)
- DOE Contract Number:
- W-31-109-ENG-38
- OSTI ID:
- 5534656
- Report Number(s):
- EPRI-NSAC-11; ON: DE82004493; TRN: 82-006613
- 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 ACCIDENTS
REACTOR CORES
COMPUTER CALCULATIONS
HEAT TRANSFER
HYDRAULICS
MATHEMATICAL MODELS
OXIDATION
PRESSURE GRADIENTS
REACTOR SAFETY
TEMPERATURE GRADIENTS
THERMAL STRESSES
ACCIDENTS
CHEMICAL REACTIONS
ENERGY TRANSFER
ENRICHED URANIUM REACTORS
FLUID MECHANICS
MECHANICS
POWER REACTORS
PWR TYPE REACTORS
REACTOR COMPONENTS
REACTORS
SAFETY
STRESSES
THERMAL REACTORS
WATER COOLED REACTORS
WATER MODERATED REACTORS
220900* - Nuclear Reactor Technology- Reactor Safety
210200 - Power Reactors
Nonbreeding
Light-Water Moderated
Nonboiling Water Cooled