Critical heat flux estimation for annular channel geometry Public Deposited

http://ir.library.oregonstate.edu/concern/graduate_thesis_or_dissertations/nz8062791

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  • Critical Heat Flux (CHF) is an important safety parameter for the design of nuclear reactors. The most commonly used predictive tool for determination of CHF is a look-up table developed using tube data with an average hydraulic test diameter of 8 mm. There exist in the world today nuclear reactors whose geometry is annular, not tubular, and whose hydraulic diameter is significantly smaller than 8 mm. In addition, any sub-channel thermal hydraulic model of fuel assemblies is annular and not tubular. Comparisons were made between this predictive tool and annular correlations developed from test data. These comparisons showed the look-up table over-predicts the CHF values for annular channels, thus questioning its ability to perform correct safety evaluations. Since no better tool exists to predict CHF for annular geometry, an effort was undertaken to produce one. A database of open literature annular CHF values was created as a basis for this new tool. By compiling information from eighteen sources and requiring that the data be inner wall, unilaterally, uniformly heated with no spacers or heat transfer enhancement devices, a database of 1630 experimental values was produced. After a review of the data in the database, a new look-up table was created. A look-up table provides localized control of the prediction to overcome sparseness of data. Using Shepard's Method as the extrapolation technique, a regular mesh look-up table was produced using four main variables: pressure, quality, mass flux, and hydraulic diameter. The root mean square error of this look-up table was found to be 0.8267. However, by fixing the hydraulic diameter locations to the database values, the root mean square error was further reduced to 0.2816. This look-up table can now predict CHF values for annular channels over a wide range of fluid conditions.
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