To investigate the effect of mixing-vane shape, heat flux at departure from nucleate boiling (DNB) and pressure loss were measured. Computational fluid dynamics (CFD) was utilized to discuss the flow control. The pressure loss and the DNB tests were performed in a water and a Freon loops, respectively. Two mixing-vanes were designed to have same projection area but different inclination. The rod-bundle was 5 by 5 and 17 by 17 respectively at the water and Freon tests. The experimental results showed that the slightly inclined mixing-vane produced the same DNB heat flux as the deeply inclined mixing-vane and did smaller pressure loss than it. Pressure loss of the two mixing-vane grids was different in spite of the same projection area. The result of CFD showed a swirl flow decaying along the main stream in the axial direction. The swirl was stronger in the deeply inclined mixing-vane, however it decayed faster whereas one maintained long in the slightly inclined mixing-vane. This result suggested that the deep inclination caused a steep change in axial momentum to induce strong turbulence diffusion. This flow structure did not change the DNB heat flux because the two-phase discontinuity dominated the phenomena. This study provided a successful example of flow control in a mixing-vane grid.

A long-term flow-induced vibration and wear test was performed for a full-scale 17×17 PWR fuel mockup, and the test results were compared with numerical simulations. The flow-induced vibration on a fuel assembly or fuel rods may cause Grid-to-Rod Fretting (GTRF) and result in the leakage of fuel rods in PWRs. GTRF involves non-linear vibration of a fuel rod due to the excitation force induced by coolant flow around a fuel rod. So, the numerical simulation is performed by VITRAN (Vibration Transient Analysis Non-linear) and Computational Fluid Dynamics (CFD). VITRAN code was developed by Westinghouse to simulate fuel rod flow induced vibration and GTRF. In this paper, it was confirmed that the code can reproduce GTRF wear for NFI fuel assembly. CFD calculation is performed to obtain the axial and lateral flow velocity around the fuel rods, reflecting detailed geometries of fuel assembly components like bottom nozzle, spacer grids. The numerical simulation reasonably reproduced the vibration and wear test for NFI fuel assembly.

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.