Interpretation of the results of the CORA-33 dry core BWR test
- Oak Ridge National Lab., TN (United States)
- Kernforschungszentrum Karlsruhe (Germany)
All BWR degraded core experiments performed prior to CORA-33 were conducted under ``wet`` core degradation conditions for which water remains within the core and continuous steaming feeds metal/steam oxidation reactions on the in-core metallic surfaces. However, one dominant set of accident scenarios would occur with reduced metal oxidation under ``dry`` core degradation conditions and, prior to CORA-33, this set had been neglected experimentally. The CORA-33 experiment was designed specifically to address this dominant set of BWR ``dry`` core severe accident scenarios and to partially resolve phenomenological uncertainties concerning the behavior of relocating metallic melts draining into the lower regions of a ``dry`` BWR core. CORA-33 was conducted on October 1, 1992, in the CORA tests facility at KfK. Review of the CORA-33 data indicates that the test objectives were achieved; that is, core degradation occurred at a core heatup rate and a test section axial temperature profile that are prototypic of full-core nuclear power plant (NPP) simulations at ``dry`` core conditions. Simulations of the CORA-33 test at ORNL have required modification of existing control blade/canister materials interaction models to include the eutectic melting of the stainless steel/Zircaloy interaction products and the heat of mixing of stainless steel and Zircaloy. The timing and location of canister failure and melt intrusion into the fuel assembly appear to be adequately simulated by the ORNL models. This paper will present the results of the posttest analyses carried out at ORNL based upon the experimental data and the posttest examination of the test bundle at KfK. The implications of these results with respect to degraded core modeling and the associated safety issues are also discussed.
- Research Organization:
- Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States)
- Sponsoring Organization:
- Nuclear Regulatory Commission, Washington, DC (United States)
- DOE Contract Number:
- AC05-84OR21400
- OSTI ID:
- 10192996
- Report Number(s):
- CONF-931079-6; ON: DE94002490; TRN: 93:025678
- Resource Relation:
- Conference: 21. water reactor safety information meeting,Bethesda, MD (United States),25-27 Oct 1993; Other Information: PBD: [1993]
- Country of Publication:
- United States
- Language:
- English
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21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS
BWR TYPE REACTORS
REACTOR ACCIDENTS
REACTOR SAFETY EXPERIMENTS
TEST FACILITIES
COMPUTERIZED SIMULATION
KERNFORSCHUNGSZENTRUM KARLSRUHE
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DATA ANALYSIS
FAILURES
FUEL CANS
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REACTOR SAFETY
GAS PRODUCTION RATES
FUEL ELEMENTS
220900
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POWER REACTORS
NONBREEDING
LIGHT-WATER MODERATED
BOILING WATER COOLED