Heat transfer uncertainty analysis for transient CHF and post CHF rod bundle experiments
CHF and post CHF rod bundle heat transfer experiments are used to provide a data base which is then applied to developing heat transfer models for nuclear reactor safety analysis. The data is obtained using a prototypical rod bundle constructed from electrical fuel rod simulators. The fuel rod simulators are instrumented with thermocouples on the inside surface of the clad to measure the rod temperature during a transient. The temperature data from the rod along with the physical description of the rod and the measured power are then used in an inverse conduction code to calculate the transient surface heat flux and thus a heat transfer coefficient. In post CHF film boiling heat transfer regimes, the absolute level of the rod surface heat fluxes or heat transfer coefficients are small such that the uncertainties in the test power, physical properties of the fuel rod simulator, rod dimensions, and the measured temperature transient can become significant. In the high heat flux conditions during the nucleate boiling heat transfer regime, the temperature difference between the fluid and heater rod surface is relatively small and the location of the thermocouple within the heater rod is a main contributor to the heat transfer uncertainty. The uncertainties in fuel rod simulator properties that impact the high heat flux calculated heat transfer coefficients will be shown to be different than in the low heat transfer post CHF type experiments. This paper discusses a method which can be used to calculate the heat transfer coefficient uncertainty for both of these experiments.
- Research Organization:
- Westinghouse Electric Company, Pittsburgh, PA
- OSTI ID:
- 6816709
- Journal Information:
- Am. Soc. Mech. Eng., (Pap.); (United States), Journal Name: Am. Soc. Mech. Eng., (Pap.); (United States) Vol. 83-HT-106; ISSN ASMSA
- Country of Publication:
- United States
- Language:
- English
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BOILING
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FUEL ASSEMBLIES
FUEL ELEMENTS
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