Freon thermal hydraulic test is expected to be one of the workable methods to develop high thermal hydraulic performance PWR fuel. That is, high pressure water and high heat flux condition in PWR core can be substituted with lower pressure Freon and lower heat flux by applying appropriate fluid-to-fluid similarity and modeling parameters. Freon DNB tests and mixing tests were carried out against a 4×4 rod bundle configuration where R-134A flowed vertically upwardly. The tests were carried out at Freon thermal hydraulic test loop in Korea Atomic Energy Research Institute (KAERI). The spacer grid used in these tests was modeled on that of conventional PWR fuel, that is, square lattice grid with split type mixing vanes. Diameter of heater rod simulating PWR fuel rod is about 10.7mm and heating length is about 2000 mm. Freon mixing tests were carried out to estimate Turbulence Diffusivity Coefficient (TDC), which was normally used in conventional thermal hydraulic design of nuclear reactor. Freon CHF test results showed that parametric trends agreed with those of existing CHF data. To predict CHF of 4×4 rod bundle, subchannel analysis code Modified COBRA-3C and NFI-1 DNB correlation were applied. TDC value used in subchannel analysis was determined by fitting Freon mixing test data. NFI-1 DNB correlation was developed for predicting DNB heat flux in rod bundle configuration by using water CHF test results at HTRF test loop at Columbia University. The design of spacer grids used in KAERI Freon DNB test was similar to that used in water CHF test at HTRF. Water equivalent flow condition of this R-134A test was estimated using fluid-to-fluid similarities. NFI-1 DNB correlation was applied to this water equivalent condition to estimate water equivalent DNB heat flux. Then R-134A equivalent DNB heat flux was estimated reversely, and compared to Freon DNB test result. The test results were predicted well and applicability of NFI-1 DNB correlation and fluid-to-fluid similarities in 4×4 rod bundle is discussed.

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