Assessment of models for steam release from concrete and implications for modeling corium behavior in reactor cavities
Conference
·
OSTI ID:6827807
Models for concrete outgassing have been developed and incorporated into a developmental version of the CONTAIN code for the assessment of corium behavior in reactor cavities. The resultant code, referred to as CONTAIN/OR in order to distinguish it from the released version of CONTAIN, has the capability to model transient heat conduction and concrete outgassing in core-concrete interaction problems. This study focused on validation and assessment of the outgassing model through comparisons with other concrete response codes. In general, the model is not mechanistic; however, there are certain important processes and feedback effects that are treated rigorously. The CONTAIN outgassing model was compared against two mechanistic concrete response codes (USINT and SLAM). Gas release and temperature profile predictions for several concrete thicknesses and heating rates were performed with acceptable agreement seen in each case. The model was also applied to predict corium behavior in a reactor cavity for a hypothetical severe accident scenario. In this calculation, gases evolving from the concrete during nonablating periods fueled exothermic Zr chemical reactions in the corium. Higher corium temperatures and more concrete ablation were observed when compared with that seen when concrete outgassing was neglected. Even though this result depends somewhat upon the makeup of the corium sources and the concrete type in the cavity, it does show that concrete outgassing can be important in the modeling of corium behavior in reactor cavities. In particular, the need to expand the traditional role of CORCON from steady-state ablation to the consideration of more transient events is clearly evident as a result of this work. 5 refs., 11 figs., 1 tab
- Research Organization:
- Sandia National Labs., Albuquerque, NM (USA)
- DOE Contract Number:
- AC04-76DP00789
- OSTI ID:
- 6827807
- Report Number(s):
- SAND-88-2329C; CONF-8810155-27; ON: DE89004376
- Country of Publication:
- United States
- Language:
- English
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Related Subjects
21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS
210100 -- Power Reactors
Nonbreeding
Light-Water Moderated
Boiling Water Cooled
22 GENERAL STUDIES OF NUCLEAR REACTORS
220900* -- Nuclear Reactor Technology-- Reactor Safety
ACCIDENTS
BUILDING MATERIALS
BWR TYPE REACTORS
C CODES
CARBON COMPOUNDS
CARBON DIOXIDE
CARBON OXIDES
CHALCOGENIDES
COMPARATIVE EVALUATIONS
COMPUTER CODES
CONCRETES
CORIUM
ENERGY TRANSFER
FLUID FLOW
FLUID MECHANICS
HEAT TRANSFER
MATERIALS
MECHANICS
OXIDES
OXYGEN COMPOUNDS
PHYSICAL PROPERTIES
REACTOR ACCIDENTS
REACTOR SAFETY
REACTORS
S CODES
SAFETY
STEAM
U CODES
WATER COOLED REACTORS
WATER MODERATED REACTORS
210100 -- Power Reactors
Nonbreeding
Light-Water Moderated
Boiling Water Cooled
22 GENERAL STUDIES OF NUCLEAR REACTORS
220900* -- Nuclear Reactor Technology-- Reactor Safety
ACCIDENTS
BUILDING MATERIALS
BWR TYPE REACTORS
C CODES
CARBON COMPOUNDS
CARBON DIOXIDE
CARBON OXIDES
CHALCOGENIDES
COMPARATIVE EVALUATIONS
COMPUTER CODES
CONCRETES
CORIUM
ENERGY TRANSFER
FLUID FLOW
FLUID MECHANICS
HEAT TRANSFER
MATERIALS
MECHANICS
OXIDES
OXYGEN COMPOUNDS
PHYSICAL PROPERTIES
REACTOR ACCIDENTS
REACTOR SAFETY
REACTORS
S CODES
SAFETY
STEAM
U CODES
WATER COOLED REACTORS
WATER MODERATED REACTORS