In-vessel phenomena -- CORA
Conference
·
OSTI ID:5790353
Experiment-specific models have been employed since 1986 by Oak Ridge National Laboratory (ORNL) severe accident analysis programs for the purpose of boiling water reactor experimental planning and optimum interpretation of experimental results. The large integral tests performed to date, which start from an initial undamaged core state, have involved significantly different-from-prototypic boundary and experimental conditions because of either normal facility limitations or specific experimental constraints. These experiments (ACRR: DF-4, NRU: FLHT-6, and CORA) were designed to obtain specific phenomenological information such as the degradation and interaction of prototypic components and the effects on melt progression of control-blade materials and channel boxes. Applications of ORNL models specific to the KfK CORA-16 and CORA-17 experiments are discussed and significant findings from the experimental analyses such as the following are presented: applicability of available Zircaloy oxidation kinetics correlations; influence of cladding strain on Zircaloy oxidation; influence of spacer grids on the structural heatup; and the impact of treating the gaseous coolant as a gray interacting medium. The experiment-specific models supplement and support the systems-level accident analysis codes. They allow the analyst to accurately quantify the observed experimental phenomena and to compensate for the effect of known uncertainties. They provide a basis for the efficient development of new models for phenomena that are currently not modeled (such as material interactions). They can provide validated phenomenological models (from the results of the experiments) as candidates for incorporation in the systems-level whole-core'' codes.
- Research Organization:
- Oak Ridge National Lab., TN (USA)
- Sponsoring Organization:
- NRC; Nuclear Regulatory Commission, Washington, DC (USA)
- DOE Contract Number:
- AC05-84OR21400
- OSTI ID:
- 5790353
- Report Number(s):
- CONF-9105173-3-Extd.Abst.; ON: DE91012070
- 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
ALLOYS
BWR TYPE REACTORS
CHEMICAL REACTIONS
CONTROL ELEMENTS
CORIUM
EXPERIMENT PLANNING
FUEL ELEMENTS
MELTDOWN
OXIDATION
PLANNING
REACTOR ACCIDENTS
REACTOR COMPONENTS
REACTOR SAFETY
REACTORS
SAFETY
TEST FACILITIES
WATER COOLED REACTORS
WATER MODERATED REACTORS
ZIRCALOY
ZIRCONIUM ALLOYS
ZIRCONIUM BASE ALLOYS
210100 -- Power Reactors
Nonbreeding
Light-Water Moderated
Boiling Water Cooled
22 GENERAL STUDIES OF NUCLEAR REACTORS
220900* -- Nuclear Reactor Technology-- Reactor Safety
ACCIDENTS
ALLOYS
BWR TYPE REACTORS
CHEMICAL REACTIONS
CONTROL ELEMENTS
CORIUM
EXPERIMENT PLANNING
FUEL ELEMENTS
MELTDOWN
OXIDATION
PLANNING
REACTOR ACCIDENTS
REACTOR COMPONENTS
REACTOR SAFETY
REACTORS
SAFETY
TEST FACILITIES
WATER COOLED REACTORS
WATER MODERATED REACTORS
ZIRCALOY
ZIRCONIUM ALLOYS
ZIRCONIUM BASE ALLOYS